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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 325 records · Page 18
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

The Fast Modular Reactor (FMR) - Development Plan of a New 50 MWe Gas-cooled Fast Reactor

General Atomics Electromagnetic Systems (GA-EMS) will be developing a new 50-megawatt electric (MWe) fast modular reactor (FMR), under the Department of Energy’s (DOE’s) Advanced Reactor Demonstration Program (ARDP), Advanced Reactor Concepts 2020 (ARC-20) development pathway, that provides safe, carbon free electricity, capable of incremental capacity additions. A modular design allows it to be factory-built and assembled on-site to keep the cost of capital low, while the dry-cooling facilitates siting to complement renewables in nearly any location. GA-EMS is committed to commercialization of the proposed reactor, with a demonstration by 2030, and deployment by the mid-2030s. The ultimate goal of the design is to develop flexible and dispatchable carbon-free power source for the 2035 US electricity market. The GAEMS- led team will verify that simplified characteristics (e.g., inert helium gas coolant, pellet-loaded fuel rod, installation-free of heat sink requirements, small and passive heat removal systems) of the FMR will result in a safe, maintainable, cost-effective, distributed, nuclear energygenerating station. Three key specific project objectives for the next three years include: Development of the conceptual design of the 50- MWe FMR plant, Achievement of Technology Readiness Level (TRL) 4 for key system and component technologies through in-pile tests, out-of-pile tests, and numerical experiments; and Development of robust techno-economic analysis (TEA) and pre-application licensing approach necessary for timely demonstration and eventual commercialization.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
NASA NTRS2026

Thermodynamic Properties of Uranium-Zirconium Carbide Solid Solutions with Oxygen

Uranium-zirconium carbide solid solutions are potential candidates for fuel in space nuclear reactors. To support development of advanced robust fuels for space applications, (U y Zr 1-y )C x O z solid solutions were synthesized via carbothermic reduction with uranium, carbon, and oxygen contents raging from y=0.05 – 0.3, x=0.95 – 0.98, and z = 0.02 – 0.07. These solid solutions were subsequently characterized by X-ray diffraction and elemental combustion analysis. High-temperature oxidative drop solution calorimetry in molten sodium molybdate solvent at 1073 K was employed to determine the formation, oxidation, and mixing enthalpies of (U y Zr 1-y )C x O z solid solutions, and thereby for evaluating their thermodynamic stability. The solid solutions are enthalpically less stable than the UC and ZrC end members. They also become less stable and more reactive towards oxidation with increasing uranium content. Incorporation of oxygen into solid solutions increases their solid solution-stability and improves their oxidation resistance.

Calorimetry↗
DOE OSTI2022

Perspective on “code qualifying” new graphite grades for use in advanced nuclear reactors*

The American Society of Mechanical Engineers (ASME) publishes the Boiler and Pressure Vessel (BPV) Code, which include guidance for the safe development, construction, and operation of boilers and pressure vessels. ASME BPV Code Section III “Rules for Construction of Nuclear Facility Components” Division 5 focuses on “High Temperature Reactors”. Subsection HH, subpart A lists the different materials properties that are to be measured and how those properties change due to different environmental conditions (oxidation and irradiation damage) for a graphite to be accepted for use in a high temperature reactor core (i.e., “Code Qualified”). Currently there are no nuclear graphite grades that are “Code Qualified” (i.e., a reactor designer can select a graphite grade and build their reactor without any additional testing), which is due in part to development of new graphite grades in the last 20 years and the lack of comprehensive programs needed to produce the data for the code cases. This perspective is going to discuss the requirements, as called out in the ASME BPV Code, that are necessary to “code qualify” a nuclear graphite grade but will primarily focus on the practical and technical challenges associated with irradiation-induced property changes and how to address these to assist with getting graphite ready for use in advanced nuclear reactors. These same technical challenges can be expected to arise for other materials being developed for advanced reactor concepts.

advanced nuclear reactors↗
DOE OSTI2022

Phase II Development of the Surveillance Test Articles to Improve the Design, Fabrication, and Testing

Advanced reactors, such as the molten salt reactor (MSR), require materials that will withstand harsher environments than the materials used for lower temperature water reactors. Materials used in MSR construction need not only withstand elevated temperature, temperature cycling, and neutron radiation, but must be able to withstand the corrosive molten salt environment. Significant materials research has been driven by the materials needs of the MSR due to the harsh environment and the material data requirements to support licensing. Limited operation experience with MSRs has made this challenging work. Information is very limited on materials degradation due to irradiation, molten salt corrosion, elevated temperature, and the resulting fatigue, creep, and creep-fatigue loading during operation. While efforts are underway to better understand the effects of this harsh environment on construction materials, the Advanced Reactors Technologies (ART) Program has been working to develop materials surveillance test articles that could be used in a materials surveillance program and allow for the collection of information on the materials degradation during plant operation and could support timely licensing of these advanced reactors.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Advanced Reactors Spent Fuel and Waste Streams Disposition Strategies

This report describes research and development (R&D) activities conducted during Fiscal Year 2023 (FY23) in the Advanced Fuels and Advanced Reactor Waste Streams Strategies work package in the Spent Fuel Waste Science and Technology (SFWST) Campaign supported by the United States (U.S.) Department of Energy (DOE). This report is focused on evaluating and cataloguing Advanced Reactor Spent Nuclear Fuel (AR SNF) and Advanced Reactor Waste Streams (ARWS) and creating Back-end Nuclear Fuel Cycle (BENFC) strategies for their disposition. The R&D team for this report is comprised of researchers from Sandia National Laboratories and Enviro Nuclear Services, LLC.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Advanced reactors spent fuel and waste streams disposition strategies

This report describes research and development (R&D) activities conducted during Fiscal Year 2023 (FY23) in the Advanced Fuels and Advanced Reactor Waste Streams Strategies work package in the Spent Fuel Waste Science and Technology (SFWST) Campaign supported by the United States (U.S.) Department of Energy (DOE). This report is focused on evaluating and cataloguing Advanced Reactor Spent Nuclear Fuel (AR SNF) and Advanced Reactor Waste Streams (ARWS) and creating Back-end Nuclear Fuel Cycle (BENFC) strategies for their disposition. The R&D team for this report is comprised of researchers from Sandia National Laboratories and Enviro Nuclear Services, LLC.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

Analysis of buffer-IPyC separation in TRISO fuel particles

During High Temperature Gas-cooled Reactor (HTGR) operation, and due to the neutron irradiation in the reactor core, damage of the nuclear fuel coating layers occurs. The mechanism of damage formation in the TRISO fuel is explored by the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program, in which the debonding process between coating layers was also investigated. The purpose of this paper is to report simulation results for two models. Firstly, the debonding restricted model, where no gap formation between buffer and IPyC layers is permitted. Secondly, a debonding enabled model, where the gap between those layers is created. The simulations were performed with the Bison code. The inputs of the simulated models are based on data from the AGR-1 experiment. Further, the research included simulations on spherical and aspherical fuel types. The simulations match results obtained by the AGR-1 experiment, which as such shows that the Bison code is a good computational method for simulating the behavior of the gap between buffer and IPyC layers in TRISO fuel. Based on the irradiation experiments, and the Bison simulations, it was concluded that the most common scenario is a gap formation along the buffer-IPyC interface, while the least possible scenario is the situation where there is no gap formation at the buffer-IPyC junction. The computational results confirmed that the sphericity of the fuel influences the thickness of the gap that occurs at the buffer-IPyC junction, in a way that with increasing aspect ratio the gap thickness increases. In addition, the fuel sphericity does not influence the Weibull failure probability. The results obtained for spherical and aspherical fuel are nearly identical.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Design considerations for high entropy alloys in advanced nuclear applications

The demanding operating environments of advanced nuclear reactors require the development of new nuclear materials that can withstand their increased physical, chemical, thermal, and radiation-related challenges. High-entropy alloys (HEAs) have shown often-impressive mechanical, thermomechanical, and corrosion-resistant properties, and offer a massive, unexplored compositional space which allows for the targeted development of application-specific materials. Furthermore, although still in a nascent stage, research has shown that HEAs may exhibit unique irradiation tolerance, including reduced defect production and resistance to irradiation-induced swelling and hardening. The mechanisms behind this increased tolerance are not yet well-understood, although the HEA-specific attributes of a complex energy landscape, reduced thermal conductivity, and shift in defect migration energies and pathways provide promising explanations. Here this work assesses the current and future challenges faced by structural nuclear materials, identifying the specific applications in which HEAs may provide a competitive advantage compared to industry-standard materials with the aid of Ashby material selection maps. Considerations are provided for the design of future nuclear HEAs, including calculations of nuclear-relevant properties to assist in the initial down-selection of elements depending on application requirements (e.g., low neutron capture for in-core applications), narrowing the existing compositional space of HEAs to a manageable scope.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Safety Testing and Destructive Examination of AGR-2 UO 2 Compact 3-1-1

Post-irradiation examination and elevated-temperature safety testing are being performed on compacts from the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program’s second irradiation experiment (AGR-2). The compacts in the AGR-2 irradiation experiment held either tristructural isotropic (TRISO)-coated particles containing uranium oxide fuel kernels (UO 2 ) or TRISO-coated particles containing fuel kernels with both uranium carbide and uranium oxide phases (UCO). In UO 2 TRISO particles, oxygen released by uranium fission can react with the surrounding carbon in the buffer layer to form carbon monoxide (CO). Excess CO can lead to various irradiation performance issues under certain operating conditions, such as pressure-induced fracture, kernel migration, and silicon carbide (SiC) corrosion. In UCO TRISO particles, CO formation is reduced because the chemical potential for oxidation of uranium carbide is lower than for oxidation of carbon.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Reviewing MACCS Capabilities for Assessing Tritium Releases to the Environment

Tritium has a unique physical and chemical behavior which causes it to be highly mobile in the environment. As it behaves similarly to hydrogen in the environment, it may also be readily incorporated into the water cycle and other biological processes. These factors and other environmental transformations may also cause the oxidation of an elemental tritium release, resulting in a multiple order of magnitude increase in dose coefficient and radiotoxicity. While source term development and understanding for advanced reactors are still underway, tritium may be a radionuclide of interest. It is thus important to understand how tritium moves through the environment and how the MACCS accident consequence code handles acute tritium releases in an accident scenario. Additionally, existing tritium models may have functionalities that could inform updates to MACCS to handle tritium. In this report tritium transport is reviewed and existing tritium models are summarized in view of potential updates to MACCS.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Safety Testing of AGR-5/6/7 Compacts 2-2-2 and 2-2-4

Compacts 2-2-2 and 2-2-4 from the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program’s final irradiation campaign (AGR-5/6/7) were each separately subjected to a thermal soak at 1600°C for 300 h under flowing helium to simulated conditions experienced during a postulated high-temperature gas-cooled reactor (HTGR) depressurization conduction cooldown event. The safety testing of the fuel compacts’ performance under accident conditions was conducted in the Oak Ridge National Laboratory (ORNL) Core Conduction Cooldown Test Facility (CCCTF), which utilizes a stand-alone hot cell at the Irradiated Fuels Examination Laboratory (IFEL). The CCCTF heats fuel to temperatures up to 1900°C in a non-oxidizing environment while continuously monitoring the sweep gas for radioactive 85 Kr to detect release levels associated with hermetic failure of the tristructural-isotropic (TRISO) coating surrounding each fuel kernel. In addition, certain metallic radionuclides that escape the fuel compact are collected on deposition cups that are periodically exchanged with a new cup to obtain information on the overall retention behavior of the TRISO coating layers. Because cesium can diffuse through intact pyrocarbon layers, abnormal degradation of the silicon carbide (SiC) layer in the absence of holistic TRISO coating failure is indicated by release of 134 Cs at levels equivalent to an individual particle inventory in the absence of significant 85 Kr release (Hunn et al. 2014). Additional description of the CCCTF system is provided in Appendix A.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Fabrication and Characterization of Single-Particle Compacts

The programmatic objective of the advanced low enriched uranium (aLEU) fuel Project within the Office of Nonproliferation Research and Development (NA-22) is to advance the development of nuclear reactor designs, fuel materials, and fabrication technologies capable of meeting a number of challenging customer requirements, including reactor stability, extended lifetime, and power density without refueling, while using only low-enriched uranium (i.e., less than 20% enrichment of 235 U). The tandem requirements of delivering sufficient power density and maintaining fuel performance over the course of a multidecade fuel lifetime tightly constrain the field of potential fuel options. Monolithic fuel forms (e.g., fuel pellets) provide maximized density of fissile material, but they may face challenges in maintaining their structure and performance over the long reactor lifetimes envisioned in this program. Dispersion fuel forms with an inert matrix provide enhanced long-term stability but sacrifice fissile material density, and they may not provide sufficient power density over the reactor lifetime with the limit of low enrichment.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Physical & Cyber Security Modeling Interfacing Through Dante and ARCADE

Physical security is increasingly facing new threats from cyber attackers, for which there is little research in the way of characterizing this threat. This report discusses the efforts to combine cyber and physical security modeling tools to investigate this novel combinatorial threat space. To accomplish this, the Dante force-on-force modeling and simulation software and the Advanced Reactor Cyber Analysis and Development Environment (ARCADE) were integrated. Dante provides a 3D environment which models the physical world, while ARCADE provides the cyber and control systems world.

42 ENGINEERING↗
DOE OSTI2023

Thermomechanical Properties of Neutron Irradiated Al 3 Hf-Al Thermal Neutron Absorber Materials

A thermal neutron absorber material composed of Al 3 Hf particles in an aluminum matrix is under development for the Advanced Test Reactor. This metal matrix composite was fabricated via hot pressing of high-purity aluminum and micrometer-size Al 3 Hf powders at volume fractions of 20.0, 28.4, and 36.5%. Room temperature tensile and hardness testing of unirradiated specimens revealed a linear relationship between volume fraction and strength, while the tensile data showed a strong decrease in elongation between the 20 and 36.5% volume fraction materials. Tensile tests conducted at 200 °C on unirradiated material revealed similar trends. Evaluations were then conducted on specimens irradiated at 66 to 75 °C to four dose levels ranging from approximately 1 to 4 dpa. Tensile properties exhibited the typical increase in strength and decrease in ductility with dose that are common for metallic materials irradiated at ≤0.4T m . Hardness also increased with neutron dose. The difference in strength between the three different volume fraction materials was roughly constant as the dose increased. Nanoindentation measurements of Al 3 Hf particles in the 28.4 vol% material showed the expected trend of increased hardness with irradiation dose. Transmission electron microscopy revealed oxygen at the interface between the Al 3 Hf particles and aluminum matrix in the irradiated material. Scanning electron microscopy of the exterior surface of tensile tested specimens revealed that deformation of the material occurs via plastic deformation of the Al matrix, cracking of the Al 3 Hf particles, and to a lesser extent, tearing of the matrix away from the particles. The fracture surface of an irradiated 28.4 vol% specimen showed failure by brittle fracture in the particles and ductile tearing of the aluminum matrix with no loss of cohesion between the particles and matrix. The coefficient of thermal expansion decreased upon irradiation, with a maximum change of –6.3% for the annealed irradiated 36.5 vol% specimen.

36 MATERIALS SCIENCE↗
DOE OSTI2021

AGR-5/6/7 Fuel Fabrication Report

The U.S. Department of Energy Office of Nuclear Energy (DOE NE) and the Idaho National Laboratory (INL) Advanced Reactor Technologies (ART) Advanced Gas Reactor (AGR) Fuel Development and Qualification program (referred to as AGR Fuel program hereafter) are pursuing qualification of tristructural isotropic (TRISO) coated particle fuel for use in high temperature gas cooled reactors (HTGRs). The AGR Fuel program was established to provide a fuel qualification data set in support of the licensing and operation of an HTGR. BWX Technologies Nuclear Operations Group (BWXT-NOG) was subcontracted to fabricate the fuel for the AGR program. Several investments and innovations were realized in preparation to fabricate fuel for the AGR-5/6/7 irradiation experiments that brought fuel fabrication fully out of the laboratory and into engineering-scale operations. These included: • Increased the kernel fabrication line capacity and uniformity • Upgraded ancillary support equipment and processes for the tristructural isotropic (TRISO) coating furnace • Demonstrated efficient production of the matrix precursor powder by dry jet milling of co mingled components • Demonstrated an engineering-scale method for quick and efficient overcoating TRISO particles with the matrix precursor • Demonstrated an automated, multi cavity compacting system with a volumetric feed system • Demonstrated a combined-cycle thermal treatment furnace These changes increased production rates of some of these processes by an order of magnitude or more while eliminating the use of flammable solvents, multiple grinding and sorting operations, and the weighing out of individual die charges. Three fuel kernel lots were fabricated for production of the fuel for AGR-5/6/7. The initial lot (J52R-16-39316) was certified to fuel specifications but was not used because the kernels had a high fraction of internal fissures that caused an unacceptable fraction of the kernels to fragment when charged to the coating furnace where the TRISO coating would be deposited. Fragmented kernels increased the dispersed uranium in the particles and produced a worrisome fraction of dimpled particles with an elevated probability of in-pile failure. After some efforts to identify the cause of the fissure formation, two additional lots were produced with much lower fissure fractions, J52R-16-69317 and 69318. The latter kernel lot was a backup to the first and was not needed. Multiple kernel batches were composited to form each of the lots so as to simulate a commercial-scale operation where kernel batches would also be composited. Multiple TRISO coating runs were performed and the product characterized so that several could be composited into a TRISO lot. TRISO lot J52R-16-98005 conformed to all fuel specifications except the mean outer pyrocarbon (OPyC) layer thickness was thinner than specified. Furthermore, it was determined that the TRISO lot had a dispersed uranium fraction (DUF) that might result in the compacts not meeting the DUF specification. A review of the role of the OPyC layer and consequences of the DUF by the Technical Coordination Team and INL concluded that the fuel was acceptable for use in the AGR-5/6/7 irradiation experiment. The TRISO particles were overcoated with the matrix precursor that had been produced in a jet-milling operation. The overcoating was performed in equipment originally designed to coat pharmaceuticals. The overcoating process performed well; producing highly spherical and uniform overcoats requiring little upgrading and no recycle or rework. TRISO particles were overcoated with the matrix precursor to achieve nominal volumetric packing fractions (PFs) of TRISO particles of 25% and 40% for the irradiation experiments. The 40% PF compacts occupy the first and fifth test capsule in the test train while the inner three capsules are loaded with 25% PF compacts. The resinated graphite matrix precursor powder was a derivative of the German A3-27 matrix formulation, which differs from previous AGR irradiation campaigns that used an A3-3 formulation. Jet milling of the matrix powder precursor produced a finer mean graphite particle size than the milling operations used for the A3-3 matrix powder precursor. Changes made in the matrix formula and equipment yielded compacts with significantly higher matrix density than was attained in previous AGR irradiation campaigns. The changes in the matrix formulation and the means of milling the powders also complicated resolution of the three fuel compact defect fractions, DUF, exposed kernel fraction (EKF), and the silicon carbide defect fraction (SDF). Characterization data from BWXT-NOG had some anomalous results, so samples of the fuel compacts and overcoated TRISO particles were also analyzed by Oak Ridge National Laboratory (ORNL) to ensure that the defect fractions were accurately characterized.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗