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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 91 records · Page 5
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Control rods for light water reactors

A control rod for a nuclear fuel assembly is described herein that includes a neutron absorbing material having a melting point greater than 1500° C. that does not form a eutectic with a melting point less than 1500° C., and may further include a cladding material having a melting point greater than 1500° C. The cladding material is selected from the group consisting of silicon carbide, zirconium, a zirconium alloy, tungsten, and molybdenum. The absorbing material is selected from the group consisting of Gd2O3, Ir, B4C, Re, and Hf. The metal cladding or the absorbing material may be coated with an anti-oxidation coating of Cr with or without a Nb intermediate layer.

Lahoda, Edward J.↗
DOE OSTI2022

Update on NNL TSL Validation Efforts [Slides]

This presentation covers how a new sub-thermal neutron transmission capability at RPI LINAC is now in use to help validate TSL evaluations. Additionally, NNL has contributed multiple material evaluations to ENDF/B-VIII.1. Lastly, this presentation covers new methods to improve TSL physics and accuracy are being formulated.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
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↗
NASA NTRS2001

Fractographic Analysis of HfB2-SiC and ZrB2-SiC Composites

Hafnium diboride-silicon carbide and zirconium diboride-silicon carbide composites are potential materials for high temperature leading edge applications on reusable launch vehicles. In order to establish material constants necessary for evaluation of in-situ fracture, bars fractured in four point flexure were examined using fractographic principles. The fracture toughness was determined from measurements of the critical crack sizes and the strength values, and the crack branching constants were established to use in forensic fractography of materials for future flight applications. The fracture toughnesses range from about 13 MPam (sup 1/2) at room temperature to about 6 MPam (sup 1/2) at 1400 C for ZrB2-SiC composites and from about 11 MPam (sup 1/2) at room temperature to about 4 MPam (sup 1/2) at 1400 C for HfB2-SiC composites.

Mecholsky, J.J., Jr.↗
DOE OSTI2021

Microstructural characterization of as-fabricated monolithic plates with boron carbide, aluminum boride, and zirconium boride burnable absorbers

The use of burnable absorbers can be beneficial for nuclear reactors by extending the fuel’s operational cycle, providing additional criticality control, and flattening the power profile. In this work, three burnable absorber materials (boron carbide, aluminum boride, and zirconium boride) embedded in aluminum have been fabricated into foils and clad in AA-6061 for potential use in high performance research reactors. The as-fabricated boron-containing phases were determined using transmission electron microscopy to be AlB2, B4C, and ZrB2. TEM also revealed weak bonding at the B4C-matrix interface. SEM showed a relatively uniform spatial distribution of boron-containing phases for all the candidate materials. Higher porosity was observed in the foil containing ZrB2 in its as-rolled condition. The porosity in the ZrB2 foil was reduced by hot isostatic pressing. The size and shape distributions of the boron-containing phases were analyzed on the criteria of cross-sectional area, perimeter, roundness, circularity, and aspect ratio. A method of converting the 2D burnable absorber dispersoids seen in cross-sectional microscopy images into 3D volumes was derived using both spherical and ellipsoidal geometry models. The difference in calculated burnable absorber dispersoid average volume between the two models ranges from 20% to 100%, which could impact burnable absorber burnout rates due to differences in neutron self-shielding.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1952

Friction and Surface Damage of Several Corrosion-resistant Materials

Friction and surface damage of several materials that are resistant to corrosion due to liquid metals was studied in air. The values of kinetic friction coefficient at low sliding velocities and photomicrographs of surface damage were obtained. Appreciable surface damage was evident for all materials tested. The friction coefficients for the combinations of steel, stainless steel, and monel sliding against steel, stainless steel, nickel, Iconel, and Nichrome ranged from 0.55 for the monel-Inconel combination to 0.97 for the stainless-steel-nickel combination; for steel, stainless steel, monel, and tungsten carbide against zirconium, the friction coefficient was approximately 0.47. Lower coefficients of friction (0.20 to 0.60) and negligible surface failure at light loads were obtained with tungsten carbide when used in combination with various plate materials.

ONTACT SURFACES, SLIDING↗
DOE OSTI2025

Ta–Zr carbides: Synthesis advances via carbothermal reduction and defect evolution observed through transmission electron microscopy ion irradiation

The thermodynamic stability of six distinct compositions within the Zr—Ta—C ternary system is investigated in this study, marking the first report of their synthesis through carbothermic reduction in vacuum. A prolonged annealing process at 2200°C enabled high densification and phase equilibrium. Detailed phase identification and microstructural characterization through microscopy and X-ray diffraction techniques revealed clear compositional trends and stable phase formations. Two compositions ((Ta 0.2 Zr 0.8 )C 0.6 and (Ta 0.5 Zr 0.5 )C 1 ) were selected for ion irradiation experiments using 200 keV Kr + at 600°C—representing the first-ever irradiation study on the Zr—Ta—C system. The findings indicated defect accumulation and nanoscale cavity formation without any evidence of amorphization, highlighting the system's structural stability under irradiation. Together, the synthesis and irradiation results provide a basis for further investigation of the system and suggest its relevance for applications under extreme environments.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Diffusion study of uranium mononitride/zirconium carbide composite for space nuclear propulsion

Here, the next generation of space exploration will require extensive developments in rocket technology. Space nuclear propulsion is of interest due to its high fuel density and power, however, it also has high temperature and stability demands. This study examines a uranium nitride/zirconium carbide ceramic-ceramic particle composite as a potential fuel for these missions. Based on the constituent properties, this fuel composite is expected to be highly thermally efficient and resistant to the hot hydrogen propellant. One of the main concerns with this composite is unknown diffusion behavior between UN and ZrC over time. Diffusion couples and composites were used to determine how these constituents will behave when left in contact at high temperatures. UN and ZrC were found to have limited but observable diffusion at the phase boundary. The resulting phase is a UZr(CN) quaternary phase. UZr(CN) has been examined for other high temperature and gas nuclear reactors with preliminary success; however, there lacks sufficient data to fully qualify this phase. Based on the limited information available, the resultant quaternary phase could be deemed acceptable and even provide better adhesion for the fuel particles (UN) to the matrix (ZrC).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

The α-WC(0001) Surface as a Hydrogen Sponge: A First Principle Study of H 2 Dissociation and Formation of Low and High Coverages

Tungsten carbide (WC) displays a Pt-like behavior in catalysis, applied in hydrogenation processes. Numerous theoretical studies have modeled the behavior and use of adsorbed hydrogen without obtaining a general picture, missing basic links between H 2 dissociation and generation of high surface coverage (Θ H >0.5 ML). Here, in this study, the capability of C- and W-terminations of the α-WC(0001) surface is analyzed to dissociate several H 2 molecules to produce coverages, Θ H , ranging from low to very high values (0.13<Θ H <2.00 ML). Density functional theory and an ab initio atomistic thermodynamic were used to achieve the conditions for H 2 dissociation. The WC–C surface has higher capacity to dissociate H 2 molecules than WC–W. However, both surfaces can reach full surface coverage, Θ H =1 ML, at mild ambient conditions, T=300 K and P=1 atm, and even up to 500 K at low and high pressures. The H-adatoms on WC–W are more labile than on WC–C. The binding of adsorbates is hindered at high Θ H , implying a need to modulate Θ H according to the application. The results give the basis to understand the capabilities of WC-based catalysts in hydrogenation-related reactions, with the advantage of WC being a hydrogen reservoir at mild practical catalytic conditions.

03 NATURAL GAS↗
DOE OSTI2022

Capability to Process and Characterize Uranium-Zirconium (U-Zr) and Uranium-Zirconium-Plutonium (U-Zr-Pu) Alloys

This work investigated co-reduction of anhydrous compounds of uranium, zirconium, and plutonium to produce uranium rich ternary nuclear fuel alloys. Metallothermic co-reduction is a novel method to produce all-metal nuclear fuels. Metal fuels offer thermal, fissility, compatibility, and security benefits over oxide fuels. Alloys of uranium–10% zirconium with plutonium contents of 0%, 2.5%, 5%, and 10% were produced, with yields of 60–85% of theoretical values in a traditional calciothermic bomb reduction apparatus. Microstructural analysis indicated transformation of uranium phase from gamma to beta and then alpha, in alternating lamellar plates typical of alpha phase uranium and delta phase uranium-zirconium, with zirconium and carbides at prior grain boundaries. Some of the analyses were inconclusive in their results and therefore require additional testing. Successful ternary co-reduction would simplify production of homogeneous feedstock and thereby streamline manufacture of homogeneous ternary metallic fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS2021

Microstructural Observations on Advanced Multilayer Protective Coatings for Nuclear Thermal Propulsion

The present research discusses the microstructures of a new multilayered coating concept proposed to solve the “midrange corrosion” problem observed in NbC and ZrC-coated graphite (Gr) nuclear fuels during the NERVA/Rover programs. The concept envisions designing a compliant multilayered coatings architecture to accommodate the thermal expansion mismatch strains between the ZrC outer coating and the Gr/(U,Zr)C fuel matrix along with using a diffusion barrier to minimize carbon and hydrogen diffusion, respectively. Multilayered coatings were deposited on Gr substrates by chemical vapor deposition (CVD). Graphite disks, as well as the inner channels of a 19-hole hexagonal rod 152.4 mm long, were coated by CVD in proof-of concept process demonstration studies. Coated disk specimens were thermally-cycled between ambient and 1900 K for as long as 20 cycles with a 1h hold time at the temperature using both stepped and non-interrupted thermal cycling methods. Detailed cross-sectional microstructural analyses were conducted on the thermally-cycled coated disks. These results showed that the coatings were intact after 20 thermal cycles at 1900 K although the bonding of the overlay Mo coating appeared to be influenced by the oxygen content in the layer. Seven transverse sections were cut along the length from the coated hexagonal rod for microstructural studies. All the nineteen channels were observed to have been coated in all the sections. This paper reports the results of detailed microstructural analyses, compositional line scans and x-ray dot maps conducted on three coated channels from a mid-section of the rod. The measured values of carbon (C) were observed to be exceptionally high in all the specimens, which were attributed to hydrocarbon contamination of the specimen surfaces. It was observed that the oxygen (O) levels were high in all the coating layers even after approximately correcting for the observed high values of C. The source of O contamination of the CVD reactor system was attributed to either the presence of oxychloride impurities in the precursor salts or a pin hole leak in the coating system. It is concluded that high values of O result in poor bonding of the Mo overlay to the ZrC layer. These trials have established that it is possible to deposit multilayered coatings by CVD in 19-hole hexagonal rods 152.4 mm long similar to the NERVA fuel rod design although the quality of the coatings and the uniformity of their thicknesses require further improvement.

Nuclear thermal propulsion, multilayer coatings, m↗
NASA NTRS1971

Vaporization of the group IVB metal carbides

High temperature vaporization of titanium, zirconium, hafnium and thorium carbides by Knudsen effusion mass spectrometry, measuring ion intensities, formation enthalpies and dissociation energies

Kohl, F. J.↗
NASA NTRS1985

Vapor pressure and evaporation rate of certain heat-resistant compounds in a vacuum at high temperatures

The vapor pressure and evaporation rate of borides of titanium, zirconium, and chrome; and of strontium and carbides of titanium, zirconium, and chrome, molybdenum silicide; and nitrides of titanium, niobium, and tantalum in a vacuum were studied. It is concluded that all subject compounds evaporate by molecular structures except AlB sub 12' which dissociates, losing the aluminum.

Bolgar, A. S.↗
NASA NTRSDate not supplied

3D Printed Ceramic Reinforced Polymer Composites Microstructure and Mechanical Properties

In Fused Filament Fabrication (FFF) 3D printing a part is made by extruding material layer by layer. Benefits of FFF are the ability to print complex shapes quickly and at a lower cost compared to other manufacturing methods. In this work, filaments composed of either silicon carbide (SiC) or zirconium silicate ceramic particulates in a polylactic-acid (PLA) matrix were used to print specimens for microstructural analysis and mechanical testing. Pure PLA samples were also printed and tested as a baseline for comparison. A Design of Experiments approach was used to test the interactions between three different parameters and their effect on ultimate strength, Young’s modulus, and yield strength. For compression test specimens, these parameters were infill percentage, number of shells, and print orientation angle. For tensile test specimens, the parameters were layer height, number of shells, and infill orientation angle. Pre- and post-compression test samples were evaluated by scanning electron microscopy for microstructural analysis. Infill percentage had the greatest effect on mechanical properties while number of shells and print orientation angle had minor effects. When changing the layer lines from perpendicular to parallel, relative to the loading direction, SiC-PLA had higher mechanical properties while pure PLA properties had decreased, and zirconium silicate-PLA had mix results.

Fused filament fabrication↗