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Results for “Uranium-zirconium”

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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11 records
All sourcesNASA NTRSDOE OSTI
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 OSTI2025

Novel Synthesis and Characterization of Uranium-Zirconium Carbonitride by Direct Casting

UZrCN has exhibited thermophysical properties beneficial to high temperature reactor applications such as nuclear thermal rockets. The present work investigates a novel liquid phase synthesis method involving the admission of nitrogen gas during arc melting of uranium, zirconium, and carbon. Initial microstructural examinations using SEM indicated that the samples remain heterogenous with zirconium-rich cores in a uranium-rich matrix. Heterogeneity resulted from large differences in melting temperature of the major constituents and rapid solidification. Additional EDS, combustion analysis, and inert gas fusion analysis proved that the core regions have both uranium and zirconium and establish that carbon and nitrogen are retained during the fabrication process. XRD clarifies that the light elements did not form compounds with the uranium matrix but rather incorporated into the core region forming a sub-stoichiometric UZrCN. Heat treatment performed on equimolar U-Zr-C resulted in an increase in the homogenous phase present with an XRD pattern reflecting a ternary UZrCN despite small amounts of segregation in the final button.

36 - MATERIALS SCIENCE↗
DOE OSTI2025

On Uranium-Zirconium Carbonitride Formation by Direct Casting - A Novel Synthesis Study

UZrCN has exhibited thermophysical properties beneficial to high temperature reactor applications such as nuclear thermal rockets. The present work investigates a novel liquid phase synthesis method involving the admission of nitrogen gas during arc melting of uranium, zirconium, and carbon. Initial microstructural examinations using SEM indicated that the samples remain heterogenous with zirconium-rich cores in a uranium-rich matrix. Heterogeneity resulted from large differences in melting temperature of the major constituents and rapid solidification. Additional EDS, combustion analysis, and inert gas fusion analysis proved that the core regions have both uranium and zirconium and establish that carbon and nitrogen are retained during the fabrication process. XRD clarifies that the light elements did not form compounds with the uranium matrix but rather incorporated into the core region forming a sub-stoichiometric UZrCN. Heat treatment performed on equimolar U-Zr-C resulted in an increase in the homogenous phase present with an XRD pattern reflecting a ternary UZrCN despite small amounts of segregation in the final button.

36 - MATERIALS SCIENCE↗
DOE OSTI2026

A Multiphysics Evaluation of Annular Uranium-Zirconium Metallic Fuels [Poster]

This study examines the performance of U-10Zr annular metallic fuel rodlets which were experimentally evaluated as part of the Advanced Fuels Campaign (AFC). The AFC mission is to develop novel fuel technologies and facilitate the implementation of those technologies by industry partners. A key objective is to improve steady-state and transient performance over current fuel types. The experiments of interest in this study included annular metallic U-Zr fuel rodlets within HT-9 cladding which were placed in SS-316 capsules and inserted in the Advanced Test Reactor (ATR). Certain mechanical and thermal conditions cannot be directly evaluated through experiments and fuel performance modeling is used to shed light on this evolution over time. In this study, BISON Multiphysics simulations are leveraged to investigate the state of the fuel system throughout and after the experimental conditions.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Density functional theory calculations of the mixing enthalpy of ternary uranium carbide compounds

The high melting point of uranium-zirconium carbides (U,Zr)C makes them an ideal fuel for nuclear thermal propulsion (NTP) reactors. Gaps remain in the current understanding of the U-Zr-C system due to the difficulty of conducting thermodynamic experiments at NTP operation conditions. Density functional theory calculations using the Hubbard U model (DFT+U) were performed using orbital matrix occupation (OMC) to obtain the mixing enthalpy for UC and ZrC for (U,Zr)C ternary compounds. Similarly, DFT+U calculations were also carried out for the (U,Nb)C and (U,Ta)C systems. In conclusion, the DFT results are envisioned to be used in thermodynamic assessments of the uranium carbide systems based on the CALPHAD approach to supplement the lack of experimental data for the mixing thermodynamics.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Material Characterization Report of EBR II Mark IV Fuel Element

The focus of this report is the characterization of an unirradiated as-built fuel element from the Experimental Breeder Reactor II (EBR-II) Mark IV driver fuel design. The fuel in this element is a metallic uranium-zirconium alloy, and the cladding material is HT9 stainless steel. Both the metallic fuel and cladding materials are characterized, including chemical composition, grain size, morphology, phase composition, precipitate composition, crystallography, density, and hardness.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1971

Compatibility of buffered uranium carbides with tungsten.

Results of compatibility tests between tungsten and hyperstoichiometric uranium carbide alloys run at 1800 C for 1000 and 2500 hours. These tests compared tungsten-buffered uranium carbide with tungsten-buffered uranium-zirconium carbide. The zirconium carbide addition appeared to widen the homogeneity range of the uranium carbide, making additional carbon available for reaction. Reaction layers could be formed by either of two diffusion paths, one producing UWC2, while the second resulted in the formation of W2C. UWC2 acts as a diffusion barrier for carbon and slows the growth of the reaction layer with time, while carbon diffusion is relatively rapid in W2C, allowing equilibrium to be reached in less than 2500 hours at a temperature of 1800 C.

Phillips, W. M.↗
NASA NTRS1972

Preliminary design of reactor power systems for the manned space base.

The results of design integration studies of uranium-zirconium hydride (UZr-Hx) reactor power systems for the NASA space base study program are presented. The power conversion systems investigated include the Brayton cycle, the organic Rankine cycle, the SNAP-8 mercury Rankine cycle, and thermoelectric (PbTe). The proposed space base has a 10-year life and requires 100 kWe of power. Two 50-kWe power systems with a nominal replacement life of 5 years are utilized. Parametric design data such as life, weight, radiator area, reactor outlet-temperature, reactor thermal power, and power conversion system efficiency are presented and used for the design and integration of the system with the space base.

Mckhann, G. G.↗
DOE OSTI2025

DISSOLUTION OF SURROGATE U-Zr FUEL USING ALNIFLEX CONDITIONS

Non-aluminum clad spent nuclear fuel (NASNF) stored in L Basin at the Savannah River Site (SRS) is widely varied in fuel composition, design, packaging, and physical condition. The complexity of the NASNF inventory presents significant challenges, and technology development is necessary for successful disposition. One such fuel in the inventory is metallic uranium-zirconium (U-Zr) alloy fuel, the focus of this study. Electrolytic or nitric acid only dissolution of metallic U-Zr alloy can form insoluble zirconium oxide, which results in up to 52% loss of U to insoluble solids, and can be subject to potentially uncontrolled oxidation reactions [1, 2]. The AlNiflex process was determined to be a viable dissolution flowsheet for the U-Zr alloy fuel. Under a narrow set of solution concentrations, a combination of hydrofluoric acid (HF), nitric acid (HNO3), aluminum nitrate (Al(NO3)3), and hexavalent chromium can safely dissolve U-Zr intermetallic alloys, keep Zr soluble, and not significantly corrode stainless steel (SS) vessels [3, 4.

Gogolski, Jarrod M. [Savannah River National Labor↗
DOE OSTI2025

IMPACT-01 Assembly Overview [Slides]

The Irradiated Material Properties Accelerated Characterization Test (IMPACT) series of experiments will irradiate three metallic fuel alloy specimens with embedded thermal conductivity probes in the Advanced Test Reactor (ATR). Metallic fuel alloys have long been under investigation for use in advanced reactors on account of their high thermal conductivity. Metallic fuel undergoes dramatic microstructural changes early in life due to fission gas swelling until ~2-3 at% burnup when pores interconnect, thus allowing fission gas to escape into the fuel pin plenum and swelling effectively ceases. The evolution of metallic fuel thermal conductivity during this early phase has never been successfully measured in situ. This experiment will be designed to use advances in measurement sciences to characterize how thermal transport properties evolve while in reactor. The IMPACT experiment consists of three metallic fuel rodlets with thermal conductivity probes axially centered within the fuel specimen. This presentation is an overview of the IMPACT-01 assembly process with a focus MFC fuel fabrication and experiment assembly steps.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗