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Results for “Zirconium Carbide”

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 55 records · Page 3
All sourcesNASA NTRSDOE OSTI
NASA NTRS2011

Lightweight, Ultra-High-Temperature, CMC-Lined Carbon/Carbon Structures

Carbon/carbon (C/C) is an established engineering material used extensively in aerospace. The beneficial properties of C/C include high strength, low density, and toughness. Its shortcoming is its limited usability at temperatures higher than the oxidation temperature of carbon . approximately 400 C. Ceramic matrix composites (CMCs) are used instead, but carry a weight penalty. Combining a thin laminate of CMC to a bulk structure of C/C retains all of the benefits of C/C with the high temperature oxidizing environment usability of CMCs. Ultramet demonstrated the feasibility of combining the light weight of C/C composites with the oxidation resistance of zirconium carbide (ZrC) and zirconium- silicon carbide (Zr-Si-C) CMCs in a unique system composed of a C/C primary structure with an integral CMC liner with temperature capability up to 4,200 F (.2,315 C). The system effectively bridged the gap in weight and performance between coated C/C and bulk CMCs. Fabrication was demonstrated through an innovative variant of Ultramet fs rapid, pressureless melt infiltration processing technology. The fully developed material system has strength that is comparable with that of C/C, lower density than Cf/SiC, and ultra-high-temperature oxidation stability. Application of the reinforced ceramic casing to a predominantly C/C structure creates a highly innovative material with the potential to achieve the long-sought goal of long-term, cyclic high-temperature use of C/C in an oxidizing environment. The C/C substructure provided most of the mechanical integrity, and the CMC strengths achieved appeared to be sufficient to allow the CMC to perform its primary function of protecting the C/C. Nozzle extension components were fabricated and successfully hot-fire tested. Test results showed good thermochemical and thermomechanical stability of the CMC, as well as excellent interfacial bonding between the CMC liner and the underlying C/C structure. In particular, hafnium-containing CMCs on C/C were shown to perform well at temperatures exceeding 3,500 F (.1,925 C). The melt-infiltrated CMC-lined C/C composites offered a lower density than Cf/SiC. The melt-infiltrated composites offer greater use temperature than Cf/SiC because of the more refractory ceramic matrices and the C/C substructure provides greater high-temperature strength. The progress made in this work will allow multiple high-temperature components used in oxidizing environments to take advantage of the low density and high strength of C/C combined with the high-temperature oxidation resistance of melt-infiltrated CMCs.

Wright, Matthew J.↗
NASA NTRS2009

Ceramic material suitable for repair of a space vehicle component in a microgravity and vacuum environment, method of making same, and method of repairing a space vehicle component

A precursor of a ceramic adhesive suitable for use in a vacuum, thermal, and microgravity environment. The precursor of the ceramic adhesive includes a silicon-based, preceramic polymer and at least one ceramic powder selected from the group consisting of aluminum oxide, aluminum nitride, boron carbide, boron oxide, boron nitride, hafnium boride, hafnium carbide, hafnium oxide, lithium aluminate, molybdenum silicide, niobium carbide, niobium nitride, silicon boride, silicon carbide, silicon oxide, silicon nitride, tin oxide, tantalum boride, tantalum carbide, tantalum oxide, tantalum nitride, titanium boride, titanium carbide, titanium oxide, titanium nitride, yttrium oxide, zirconium diboride, zirconium carbide, zirconium oxide, and zirconium silicate. Methods of forming the ceramic adhesive and of repairing a substrate in a vacuum and microgravity environment are also disclosed, as is a substrate repaired with the ceramic adhesive.

Riedell, James A.↗
NASA NTRS2011

Methods of repairing a substrate

A precursor of a ceramic adhesive suitable for use in a vacuum, thermal, and microgravity environment. The precursor of the ceramic adhesive includes a silicon-based, preceramic polymer and at least one ceramic powder selected from the group consisting of aluminum oxide, aluminum nitride, boron carbide, boron oxide, boron nitride, hafnium boride, hafnium carbide, hafnium oxide, lithium aluminate, molybdenum silicide, niobium carbide, niobium nitride, silicon boride, silicon carbide, silicon oxide, silicon nitride, tin oxide, tantalum boride, tantalum carbide, tantalum oxide, tantalum nitride, titanium boride, titanium carbide, titanium oxide, titanium nitride, yttrium oxide, zirconium boride, zirconium carbide, zirconium oxide, and zirconium silicate. Methods of forming the ceramic adhesive and of repairing a substrate in a vacuum and microgravity environment are also disclosed, as is a substrate repaired with the ceramic adhesive.

Riedell, James A.↗
DOE OSTI2020

Titanium-group nano-whiskers and method of production

Disclosed herein are structures comprising a titanium, zirconium, or hafnium powder particle with titanium carbide, zirconium carbide, or hafnium carbide (respectively) nano-whiskers grown directly from and anchored to the powder particle. Also disclosed are methods for fabrication of such structures, involving heating the powder particles and exposing the particles to an organic gas.

Seals, Roland D.↗
DOE OSTI2023

Titanium-group nano-whiskers and method of production

Disclosed herein are structures comprising a titanium, zirconium, or hafnium powder particle with titanium carbide, zirconium carbide, or hafnium carbide (respectively) nano-whiskers grown directly from and anchored to the powder particle. Also disclosed are methods for fabrication of such structures, involving heating the powder particles and exposing the particles to an organic gas.

Menchhofer, Paul A.↗
DOE OSTI2024

ENDF/B-VIII.1: Thermal Neutron Scattering Sublibrary

The thermal neutron scattering law (TSL) sublibrary aims to describe the interaction of incident neutrons at thermal or sub-thermal energies with different compound materials such fuels, moderators and special-purpose materials. In ENDF/B-VIII.1 there was a large number of new and updated TSL evaluations, including traditional moderators (light water, Beryllium metal, Beryllium Oxide, Calcium Hydride, plastics (Polystyrene and Lucite), graphite (reactor-grade and crystalline), anhydrous Hydrogen Fluoride, and heavy paraffinic oil); exotic moderators (Beryllium Carbide, Zirconium Hydride, Yttrium Hydride, Lithium-7 Hydride and Deuteride), FLiBe molten salt, structural materials and cladding (Silicon Carbide, Silicon Dioxide, Zirconium Carbide), fuels (Plutonium Dioxide, Uranium Carbide, Uranium metal, Uranium Nitride, Uranium Dioxide, Uranium Hydride), and special purpose materials. In ENDF/B-VIII.1 we also distribute alongside the evaluated files, a comma-separated file (CSV), named TSL_MAT_numbers.csv, which lists all evaluated files in the current release and their corresponding unique MAT number.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
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

Swift heavy ion irradiation effects in zirconium and hafnium carbides

The behavior of microcrystalline zirconium carbide (ZrC) and hafnium carbide (HfC) was studied under highly ionizing irradiation conditions at room temperature. The induced structural modifications were characterized via synchrotron-based X-ray diffraction experiments. Unit-cell expansion and buildup of microstrain were determined across a wide fluence range and linked to chemical compositions of the target material. The observed swelling resulting from irradiation with 198 MeV Xe ions in both carbide materials is characterized by two distinct mechanisms that operate within different fluence regimes. Unit-cell expansion initially proceeds by a direct-impact behavior that reaches saturation, followed at higher fluences by a second, linear swelling regime. The overall behavior, particularly the direct-impact regime, is similar for ZrC and HfC, with a more pronounced second defect accumulation process in HfC. Swelling in ZrC shows the same two distinct mechanisms upon irradiation with 198 MeV Xe ions and 946 MeV Au ions, but swelling induced by the lower-energy ions is greater across the entire fluence series. Accounting for the difference in energy deposition density between the two irradiation conditions reveals that the first swelling mechanism (direct-impact behavior) is likely related to the formation of more simple defects. In contrast, the second damaging mechanism at higher fluences (linear increase) cannot be fully explained by the induced energy density, and swelling remains somewhat higher for the low-velocity Xe irradiation. This may suggest that more complex defects and defect clusters are responsible for this swelling regime, with either their size and/or morphology modified at different energy densities.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS2023

Sintering Behaviors of ZrC, NbC, and TaC Mono-and Binary Carbides

Nuclear Thermal Propulsion (NTP) has undergone development as an alternate in-space propulsion system to traditional chemical propulsion methods since the 1950s. In an NTP system, the energy released from fission in the core is utilized as the heat source to directly heat a propellant for propulsion, rather than chemical combustion in a traditional rocket. NTP has many desirable capabilities including flexible mission launch dates and reduced transit times due to it’s capability for high specific impulse. One of the main challenges with NTP systems is the structural integrity of the fuel. The fuel form required in the reactor core must withstand temperatures above 2700 Kelvin. Ceramic-metallic matrix fuel, ceramic-ceramic matrix fuel, and solid solution carbide fuels are the three strongest candidates for the extreme environments. Solid solution carbides have the potential to exhibit promising behavior as a fuel form in an NTP system. Of the multiple refractory metal carbides of interest, here we focus on zirconium carbide (ZrC), niobium carbide (NbC), and tantalum carbide (TaC). ZrC, NbC, and TaC powders were consolidated in monocarbide (ZrC, NbC, and TaC) and bi-carbide (ZrC-NbC, ZrC-TaC, NbC-TaC) forms using spark plasma sintering (SPS). In addition to the pure endpoint carbides, the examined compositions of the various bi-carbides ranged from 25-75 mol%. The sintering temperatures, pressures, and hold times were varied to determine the ideal sintering conditions. Grain size analysis, Archimedes’ density, scanning electron microscopy (SEM), X-ray diffraction (XRD), and energy dispersion spectroscopy (EDS) were used to determine and characterize the grain size, ideal density, porosity, phase stability, and chemical composition of each sample. The data from each sample was then used to generate a Master Sintering Curve (MSC) unique to each monocarbide or bi-carbide.

nuclear thermal propulsion↗
DOE OSTI2023

A Study of a Surrogate Nuclear Thermal Propulsion Fuel Element

We report a surrogate fuel, Zirconium Carbide/Titanium Nitride has been prepared using a simple method to obtain a more uniform distribution of titanium nitride in the zirconium carbide matrix. Powders were sintered using Spark Plasma Sintering at 1750oC and 1850oC using a modified sintering profile. This resulted in a fairly uniform distribution of titanium nitride in the zirconium carbide matrix with a density of 95% of theoretical at 1850oC. Scanning electron microscopy, indentation hardness and x-ray diffraction results are given.

36 MATERIALS SCIENCE↗
NASA NTRS1972

Some observations on uranium carbide alloy/tungsten compatibility.

Results of chemical compatibility tests between both pure tungsten and thoriated tungsten run at 1800 C for up to 3300 hours with uranium carbide alloys. Alloying with zirconium carbide appeared to widen the homogeneity range of uranium carbide, making additional carbon available for reaction with the tungsten. Reaction layers were formed both by vapor phase reaction and by physical contact, producing either or both UWC2 and W2C, depending upon the phases present in the starting fuel alloy. Formation of UWC2 results in slow growth of the reaction layer with time, while W2C reaction layers grow rapidly, allowing equilibrium to be reached in less than 2500 hours at 1800 C. Neither the presence of a thermal gradient nor the presence of thoria in the tungsten clad affect the reactions observed.

Phillips, W. M.↗
DOE OSTI2022

Silicon carbide reinforced zirconium based cladding

A method for making an improved nuclear fuel cladding tube includes reinforcing a Zr alloy tube by first winding or braiding ceramic yarn directly around the tube to form a ceramic covering, then physically bonding the ceramic covering to the tube by applying a first coating selected from the group consisting of Nb, Nb alloy, Nb oxide, Cr, Cr oxide, Cr alloy, or combinations thereof, by one of a thermal deposition process or a physical deposition process to provide structural support member for the Zr tube, and optionally applying a second coating and optionally applying a third coating by one of a thermal deposition process or a physical deposition process. If the tube softens at 800° C.-1000° C., the structural support tube will reinforce the Zr alloy tube against ballooning and bursting, thereby preventing the release of fission products to the reactor coolant.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1973

Application of diffusion barriers to the refractory fibers of tungsten, columbium, carbon and aluminum oxide

A radio frequency powered ion-plating system was used to plate protective layers of refractory oxides and carbide onto high strength fiber substrates. Subsequent overplating of these combinations with nickel and titanium was made to determine the effectiveness of such barrier layers in preventing diffusion of the overcoat metal into the fibers with consequent loss of fiber strength. Four substrates, five coatings, and two metal matrix materials were employed for a total of forty material combinations. The substrates were tungsten, niobium, NASA-Hough carbon, and Tyco sapphire. The diffusion-barrier coatings were aluminum oxide, yttrium oxide, titanium carbide, tungsten carbide with 14% cobalt addition, and zirconium carbide. The metal matrix materials were IN 600 nickel and Ti 6/4 titanium. Adhesion of the coatings to all substrates was good except for the NASA-Hough carbon, where flaking off of the oxide coatings in particular was observed.

Douglas, F. C.↗