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Hot Hydrogen Testing of Uranium Nitride Cermet for Nuclear Thermal Propulsion

This study investigated the thermochemical stability and performance characteristics of uranium mononitride tungsten molybdenum cermet (UN-Mo30W) as a potential fuel material for Nuclear Thermal Propulsion (NTP) engines. Twelve monolithic cylinders were subjected to conditions simulating an active NTP engine. The tests demonstrated thermal variability within the test article and highlighted the dissociation of UN into U and N2 at elevated temperatures, affecting the microstructure and causing bonding between cermet wafers and cooling tubes. Findings indicate that while UN-Mo30W cermet exhibits some desirable properties for NTP applications, its suitability is limited by high-temperature thermochemical instability.

nuclear fuels↗

DOE OSTI2023

Formation and Consolidation of UN Nanostructures in U-Mo Fuel via Mechanical Alloying and Spark Plasma Sintering

Uranium-molybdenum (U-Mo) alloys show promise as a nuclear fuel system due to their high thermal conductivity and high fuel loading capability. However, U-Mo systems suffer from irradiation induced swelling ultimately affecting the cladding via mechanical and chemical interaction. To address these shortcomings, this thesis investigated the formation of uranium mononitride (UN) nanoparticles within a U-10Mo matrix. To promote the formation of UN, U-10Mo powders were mechanically alloyed under a high purity nitrogen cover gas (99.9995%) atmosphere. Variations were made in stainless steel milling media size, duration of milling, and number of times the milling jar was re-aerated with nitrogen gas. UN nanoparticulates were successfully formed within the U-Mo matrix and was characterized utilizing light element analysis, X-ray diffractometry, scanning and transmission-electron microscopy, electron energy loss spectroscopy, and atom probe tomography. Presence of the UN nanoparticles was found as early as 1-hour into the mechanical alloying process, and significant iron contamination found after 10-hours of milling. The U-Mo/UN powders were compacted using spark plasma sintering techniques in order to stabilize the nanostructure. U-Mo/UN compacts containing small amounts of neodymium and helium were analyzed to observe the defect migration tendencies of fission products within the fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2022

NASA CERCER Development TDM Presentation

In NASA's Space Nuclear Propulsion program there is a development effort of for a composite fuel containing zirconium carbide (ZrC) and uranium mononitride (UN), this fuel form is referred to as a CERCER (ceramic-ceramic). These series of slides were put together to showcase the primary achievements in our research over the past year to our program sponsors. This is a yearly meeting to evaluate progress.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2025

UN Synthesis, Fabrication, and Characterization and UC Oxidation Study in support of Advanced LEU Fuel Concepts

Uranium nitride and carbide-based fuels are proposed for use in advanced LEU fuel systems. This report details recent development work to study synthesis, sintering and fabrication methods for single phase UN in monolithic geometries. These samples were intended to support a variety of irradiation testing conditions by tailoring enrichment, geometry, and compositional requirements. Two methods were under evaluation for synthesizing uranium nitride powder – carbothermic reduction-nitridation (CTR-N) of UO 2 powder feedstock and hydride-dehydride/nitridation (HDN) of uranium metal feedstock. Pellets were fabricated from powder feedstocks via conventional pressing and sintering methods. As-fabricated feedstocks and pellets were evaluated via X-ray diffraction (XRD) for phase purity and sintered materials were characterized for density. Both the CTR N and the HDN methods were successfully demonstrated as an effective process for fabricating phase pure UN. Additionally, results from machining efforts for UN are detailed. This report also contains information on a study trying to elucidate the oxidation mechanisms of sintered uranium carbide samples fabricated from feedstock synthesized via an arc-melting method.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

Preliminary Conceptual Design of Nuclear Thermal Rocket Reactor Cores Using Ceramic Fuels with Beryllium or Composite Neutron Moderators

Herein several preliminary conceptual designs of nuclear thermal rocket reactor cores are presented that use tin-bonded monolithic ceramic [mononitride (UN), monocarbide (UC), and uranium dioxide (UO 2 )] fuel plates or pins with molybdenum-tungsten alloy clad. Neutron moderation is provided by a block of Be metal or composite materials using metal hydrides such as ZrH 1.6 or YH 1.6 with different matrices (MgO or Be). Mainly high-assay low-enriched uranium is considered, but highly enriched uranium is also assessed for a few configurations. Nominal core thermal power is 300 MW corresponding to about 66 kN (15 klbf) of thrust, and with minimal modifications, 500 MW may be possible (25 klbf of thrust). Depending on the configurations, the amount of 235 U needed for criticality is 30 to 90 kg, and reactor weight is 2.5 to 3.8 tonnes.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2022

Thermodynamics Modeling for Actinide Monocarbides and Mononitrides from First Principles

The high-temperature thermodynamical properties for the actinide monocarbides and mononitrides ThC, ThN, UC, UN, PuC, and PuN are calculated from first-principles electronic-structure theory. The electronic structure is modeled with density-functional theory (DFT) and is fully relativistic, including the spin-orbit interaction. Furthermore, the DFT is extended to account for orbital–orbital interactions, by means of a parameter-free orbital-polarization (OP) technique, that has proven to be essential for the 5f electrons in plutonium. Strong anharmonicity and the temperature dependence of the lattice vibrations are captured with the self-consistent ab initio lattice dynamics (SCAILD) method. The calculated free energies and heat capacities are compared to published results from quasi-harmonic (QH) theory, and experiments, where available. For the uranium and plutonium compounds, we make use of CALPHAD assessments to help evaluate the theory. Generally, our anharmonic relativistic approach compares well with both CALPHAD and experiments. For the thorium compounds, our theory is in good accord with QH modeling of the free energy at lower temperatures but for the heat capacity the comparison is less favorable.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗