Search NASA⌕ Search

SEARCH · Search NASA

Results for “Carbides”

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.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 937 records · Page 52
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Titania Composites with 2 D Transition Metal Carbides as Photocatalysts for Hydrogen Production under Visible-Light Irradiation

In the MXenes family of two-dimensional transition-metal carbides there were successful demonstrations of co-catalysts with rutile TiO 2 for visible-light-induced solar hydrogen production from water splitting. The physicochemical properties of Ti 3 C 2 T x MXene coupled with TiO 2 were investigated by a variety of characterization techniques. The effect of the Ti 3 C 2 T x loading on the photocatalytic performance of the TiO 2 /Ti 3 C 2 T x composites was elucidated. Moreover, with an optimized Ti 3 C 2 T x content of 5 wt %, the TiO 2 /Ti 3 C 2 T x composite shows a 400 % enhancement in the photocatalytic hydrogen evolution reaction compared with that of pure rutile TiO 2 . We also expanded our exploration to other MXenes (Nb 2 CT x and Ti 2 CT x ) as co-catalysts coupled with TiO 2 , and these materials also exhibited enhanced hydrogen production. These results manifest the generality of MXenes as effective co-catalysts for solar hydrogen production.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Metal–Nitrogen–Carbon Cluster–Decorated Titanium Carbide is a Durable and Inexpensive Oxygen Reduction Reaction Electrocatalyst

Clusters of nitrogen- and carbon-coordinated transition metals dispersed in a carbon matrix (e. g., Fe-N-C) have emerged as an inexpensive class of electrocatalysts for the oxygen reduction reaction (ORR). In this work, it was shown that optimizing the interaction between the nitrogen-coordinated transition metal clusters embedded in a more stable and corrosion-resistant carbide matrix yielded an ORR electrocatalyst with enhanced activity and stability compared to Fe-N-C catalysts. Utilizing first-principles calculations, an electrostatics-based descriptor of catalytic activity was identified, and nitrogen-coordinated iron (FeN 4 ) clusters embedded in a TiC matrix were predicted to be an efficient platinum-group metal (PGM)-free ORR electrocatalyst. Guided by theory, selected catalyst formulations were synthesized, and it was demonstrated that the experimentally observed trends in activity fell exactly in line with the descriptor-derived theoretical predictions. The Fe-N-TiC catalyst exhibited enhanced activity (20 %) and durability (3.5-fold improvement) compared to a traditional Fe-N-C catalyst. It was posited that the electrostatics-based descriptor provides a powerful platform for the design of active and stable PGM-free electrocatalysts and heterogenous single-atom catalysts for other electrochemical reactions.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2023

Grain size dependence of thermally induced oxidation in zirconium carbide

Here complementary analytical approaches were employed to probe the effect of grain size on thermally induced oxidation of zirconium carbide (ZrC) utilizing thermogravimetric analysis, differential scanning calorimetry, and Raman spectroscopy, as well as synchrotron-based and laboratory-based X-ray diffraction (XRD) experiments. The oxidation mechanism and phase behavior of nanocrystalline ZrC (grain size ~ 20 nm) were compared with that of the more documented microcrystalline ZrC (grain size ~ 1 µm). Synchrotron XRD at the Advanced Photon Source with a hydrothermal diamond anvil cell (HDAC) used as a sample chamber revealed that the onset of oxidation is at ~ 380 °C for microcrystalline ZrC which is in agreement with previous work. In contrast, the critical oxidation temperature was ~ 330 °C for nanocrystalline ZrC. Additional high-temperature synchrotron XRD experiments at the National Synchrotron Light Source II using a lamp furnace in combination with Raman analysis showed that tetragonal ZrO 2 forms as an initial oxidation product and transforms at higher temperatures to the monoclinic phase. Thermogravimetric analysis (TGA) coupled with differential scanning calorimetry (DSC) confirmed the X-ray results of a lower critical oxidation temperature for the nanocrystalline sample. The phase transformations in the oxide phase with associated critical temperatures were also evident in the thermodynamic data as exothermic heat events.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Cumulative fission yields of short-lived isotopes under natural-abundance-boron-carbide-moderated neutron spectrum

The availability of gamma spectroscopy data on samples containing mixed fission products at short times after irradiation is limited. Due to this limitation, data interpretation methods for gamma spectra of mixed fission product samples, where the individual fission products have not been chemically isolated from interferences, are not well-developed. The limitation is particularly pronounced for fast pooled neutron spectra because of the lack of available fast reactors in the United States. Samples containing the actinide isotopes 233, 235, 238U, 237Np, and 239Pu individually were subjected to a 2$ pulse in the Washington State University 1 MW TRIGA reactor. To achieve a fission-energy neutron spectrum, the spectrum was tailored using a natural abundance boron carbide capsule to absorb neutrons in the thermal and epithermal region of the spectrum. Our tailored neutron spectrum is unique to the WSU reactor facility, consisting of a soft fission spectrum that contains some measurable flux in the resonance region. This results in a neutron spectrum at greater than 0.1 keV with an average energy of 70 keV, similar to fast reactor spectra and approaching that of 235U fission. Unique fission product gamma spectra were collected from 4 minutes to 1 week after fission using single-crystal high purity germanium detectors. Cumulative fission product yields measured in the current work generally agree with published fast pooled fission product yield values from ENDF/B-VII, though a bias was noted for 239Pu. The present work contributes to the compilation of energy-resolved fission product yield nuclear data for nuclear forensic purposes.

Finn, Erin C.↗
DOE OSTI2019

The Evolution of Deformation-Induced Carbides during Divorced Eutectoid Transformation in GCr15 Steels

A series of rolling experiments were carried out using GCr15 steel to investigate the effect of initial microstructure on the spheroidizing process, and especially the effect of deformation-induced carbides (DIC) on the divorced eutectoid transformation (DET). The results indicated that the existence of DIC was beneficial to the spheroidizing process, the optimal initial microstructure for DET was lamellar pearlite, and DIC precipitated at grain boundaries. The average thickness of the cementite lamella and the size of DIC decreased with a decrease in rolling temperature in the critical temperature region. When the specimen was rolled at 1053 K, a small amount of DIC of size ~ 1 μm precipitated at grain boundaries. In contrast, the amount of DIC was increased when the rolling temperature was decreased to 1023 K, whereas the size of DIC was reduced to ~ 0.5 μm. Interrupted quenching experiments indicated that DIC (M{sub 7}C{sub 3}) were transformed to M{sub 3}C during the process of partial austenitization, and the evolution of spheroidization was influenced by the variation in the initial microstructure. For the specimens rolled at 1023 K, only ~ 60 min was needed to achieve the dissolution of DIC, which was the minimum, and the total time needed to achieve spheroidization was reduced to ~ 90 min.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Nanoscale Solute Partitioning and Carbide Precipitation in a Multiphase TRIP Steel Analyzed by Atom Probe Tomography

Nanoscale solute partitioning across multiple constituent phases in a 980-grade quenched and partitioned (Q&P) steel was analyzed using atom probe tomography (APT). The Q&P process was used to increase the C content in the retained austenite phase thereby improving its stability under plastic straining. Significant carbon enrichment of austenite was measured with decreased levels of C in martensite and almost depleted C content in ferrite, supporting the C partitioning mechanism in the literature. The APT analysis of retained austenite surrounded by martensite demonstrated a higher amount of C content compared with retained austenite surrounded by the ferrite phase. Lath and discrete carbide particle precipitation was also observed inside martensite colonies, tying up C and reducing the total amount of C available for austenite stabilization. In addition, the partitioning of Mn and other minor elements was quantitatively investigated by correlating APT and SEM-EBSD. Furthermore, these techniques provide a robust methodology for analyzing nanoscale compositional partitioning in multiphase steels, TRIP steels in particular, which can be used to better explain their microstructure-mechanical property relationships.

36 MATERIALS SCIENCE↗
DOE OSTI2022

3D-printed electrically conductive silicon carbide

The development of electrically conductive ceramics could achieve robust mechanical strength as well as practically high conductivity, offering applications in structural electrodes, conductors, catalyst supports, etc. However, its operating temperature is limited due to the intrinsic dense structures inevitably hindering the thermal management capability, thus resulting in a temperature-dependent electrical behavior in high-temperature environments. In this work, we report an additive manufacturing protocol through vat photopolymerization 3D printing to fabricate the architectured conductive silicon carbide (SiC) ceramics that simultaneously possess high electrical conductivity as well as low thermal conductivity, and demonstrate electric reliability under high-temperature environments above 600°C. The percolation of graphene into the ceramic scaffold establishes a uniform conductive network, exhibiting its electrical conductivity up to 1000 S m –1 . The bulk density of the 3D-printed ceramic is measured from 0.366 g cm –3 to 0.897 g cm –3 , with thermal conductivity ranging from 62 mW m –1 K –1 to 88 mW m –1 K –1 . Furthermore, the mechanical performance of conductive ceramic can be effectively reinforced by densifying the microstructures via spark plasma sintering treatment. The proposed additive manufacturing strategy widens the potential of ceramics as a structural and functional material, offering a promising pathway toward high-temperature electronics applications.

36 MATERIALS SCIENCE↗
DOE OSTI2017

Determining the minimum required uranium carbide content for HTGR UCO fuel kernels

There are three important failure mechanisms that must be controlled in high-temperature gas-cooled reactor (HTGR) fuel for certain higher burnup applications are SiC layer rupture, SiC corrosion by CO, and coating compromise from kernel migration. All are related to high CO pressures stemming from free O generated when uranium present as UO 2 fissions and the O is not subsequently bound by other elements. Furthermore, in the HTGR UCO kernel design, CO buildup from excess O is controlled by the inclusion of additional uranium in the form of a carbide, UC x . An approach for determining the minimum UC x content to ensure negligible CO formation was developed and demonstrated using CALPHAD models and the Serpent 2 reactor physics and depletion analysis tool. Our results are intended to be more accurate than previous estimates by including more nuclear and chemical factors, in particular the effect of transmutation products on the oxygen distribution as the fuel kernel composition evolves with burnup.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2025

Current activated reactive ultrafast joining (CARUJ) of silicon carbide

In this work we propose and demonstrate a novel approach for rapid fabrication of ceramic-ceramic joint assemblies, using sintered Silicon Carbide (SiC) as an initial example. Current Activated Reactive Ultrafast Joining (CARUJ) utilizes resistive heating of carbon-based materials to apply localized heat at or around the joint zone at heating rates of 101– 103 °C/min. CARUJ is used to fabricate SiC-SiC joints with a Si-SiC interface using minimal pressure (1–2 MPa) at time scales considerably shorter (several minutes opposed to hours) than conventional approaches for similar systems. The reaction of an interfacial precursor based on elemental silicon and carbon leads to in-situ SiC formation to produce a continuous and dense bond in a single step. Measured average joint strengths of roughly 15 MPa are achieved when tested in single lap offset (SLO) compressive shear. Optional additions of refractory metals such as molybdenum can be utilized to introduce secondary inclusions such as MoSi 2 within the joint interface. We further demonstrate SiC-SiC joining using Active Brazing Alloys (ABA) and proof of concept joining of tubular geometries. The localized and rapid heat application realizes a versatile material joining technique that could be extended for joining components at the plant site.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Microstructural evolution of a silicon carbide-carbon coated nanostructured ferritic alloy composite during in-situ Kr ion irradiation at 300°C 450°C

This work focuses on irradiation behaviors of a novel silicon carbide and carbon coated nanostructured ferritic alloy (SiC-C@NFA) composite for potential applications as a cladding and structural material for next generation nuclear reactors. The SiC-C@NFA samples were irradiated with 1 MeV Kr ions up to 10 dpa at 300 and 450 °C. Microstructures and defect evolution were studied in-situ at the IVEM-Tandem facility at Argonne National Laboratory. The effects of ion irradiation on various phases such as α-ferrite matrix, (Fe,Cr) 7 C 3 , and (Ti,W)C precipitates were evaluated. The α-ferrite matrix showed a continuous increase in dislocation density along with spatial ordering of dislocation loops (or loop strings) at >5 dpa. The size of the dislocation loops at 450 °C was larger than that at 300 °C. The nucleation and growth of new (Ti,W)C precipitates in α-ferrite grains were enhanced with the ion dose at 450 °C. This study provides new insight into the irradiation resistance of the SiC-C@NFA system.

(FeCr)(7)C-3↗
DOE OSTI2014

Irradiation creep of nano-powder sintered silicon carbide at low neutron fluences

The irradiation creep behavior of nano-powder sintered silicon carbide was investigated using the bend stress relaxation method under neutron irradiation up to 1.9 dpa. The creep deformation was observed at all temperatures ranging from 380 to 1180 °C mainly from the irradiation creep but with the increasing contributions from the thermal creep at higher temperatures. Microstructural observation and data analysis were performed.

36 MATERIALS SCIENCE↗
DOE OSTI2014

High dose neutron irradiations of Hi-Nicalon Type S silicon carbide composites, Part 1: Microstructural evaluations

Over the past decade, significant progress has been made in the development of silicon carbide (SiC) composites, composed of near-stoichiometric SiC fibers embedded in a crystalline SiC matrix, to the point that such materials can now be considered nuclear grade. Recent neutron irradiation studies of Hi-Nicalon Type S SiC composites showed excellent radiation response at damage levels of 30-40 dpa at temperatures of 300-800 °C. However, more recent studies of these same fiber composites irradiated to damage levels of >70 dpa at similar temperatures showed a marked decrease in ultimate flexural strength, particularly at 300 °C. Here, electron microscopy is used to analyze the microstructural evolution of these irradiated composites in order to investigate the cause of the degradation. While minimal changes were observed in Hi-Nicalon Type S SiC composites irradiated at 800 °C, substantial microstructural evolution is observed in those irradiated at 300° C. Furthermore, carbonaceous particles in the fibers grew by 25% compared to the virgin case, and severe cracking occurred at interphase layers.

36 MATERIALS SCIENCE↗
DOE OSTI2014

Silicon Carbide as a tritium permeation barrier in tungsten plasma-facing components

The control of tritium inventory is of great importance in future fusion reactors, not only from a safety standpoint but also to maximize a reactor’s efficiency. Due to the high mobility of hydrogenic species in tungsten (W) one concern is the loss of tritium from the system via permeation through the tungsten plasma-facing components (PFC). This can lead to loss of tritium through the cooling channels of the wall thereby mandating tritium monitoring and recovery methods for the cooling system of the first wall. The permeated tritium is then out of the fuel cycle and cannot contribute to energy production until it is recovered and recycled into the system. Most work on tritium permeation barriers has been focused at the breeding blanket where tritium is bred to maintain a self-sustaining tritium fuel cycle. Tritium inventory control is critical there as well to maintain a viable tritium breeding ratio, but there is also a great benefit to reducing permeation of tritium through the PFC (e.g. the first wall). This stems from the fact that the rate at which the tritium recycles between the plasma and solute state in the plasma-facing materials is much larger than any other tritium “loop” in the reactor. Here, this work demonstrates that a thin layer of silicon carbide (SiC) can be used in a W PFC to greatly reduce hydrogenic permeation while maintaining structural integrity of the PFC through rapid thermal anneal and thermal cycling tests up to 1023 K.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

High Dose Neutron Irradiation of Hi-Nicalon Type S Silicon Carbide Composites, Part 2. Mechanical and Physical Properties

Nuclear-grade silicon carbide (SiC) composite material was examined for mechanical and thermophysical properties following high-dose neutron irradiation in the High Flux Isotope Reactor at a temperature range of 573–1073 K. Likewise, the material was chemical vapor-infiltrated SiC-matrix composite with a two-dimensional satin weave Hi-Nicalon Type S SiC fiber reinforcement and a multilayered pyrocarbon/SiC interphase. Moderate (1073 K) to very severe (573 K) degradation in mechanical properties was found after irradiation to >70 dpa, whereas no evidence was found for progressive evolution in swelling and thermal conductivity. The swelling was found to recover upon annealing beyond the irradiation temperature, indicating the irradiation temperature, but only to a limited extent. Moreover, the observed strength degradation is attributed primarily to fiber damage for all irradiation temperatures, particularly a combination of severe fiber degradation and likely interphase damage at relatively low irradiation temperatures.

36 MATERIALS SCIENCE↗
DOE OSTI2016

Neutron-irradiation creep of silicon carbide materials beyond the initial transient

Irradiation creep beyond the transient regime was investigated for various silicon carbide (SiC) materials. Here, the materials examined included polycrystalline or monocrystalline high-purity SiC, nanopowder sintered SiC, highly crystalline and near-stoichiometric SiC fibers (including Hi-Nicalon Type S, Tyranno SA3, isotopically-controlled Sylramic and Sylramic-iBN fibers), and a Tyranno SA3 fiber–reinforced SiC matrix composite fabricated through a nano-infiltration transient eutectic phase process. Neutron irradiation experiments for bend stress relaxation tests were conducted at irradiation temperatures ranging from 430 to 1180 °C up to 30 dpa with initial bend stresses of up to ~1 GPa for the fibers and ~300 MPa for the other materials. Initial bend stress in the specimens continued to decrease from 1 to 30 dpa. Analysis revealed that (1) the stress exponent of irradiation creep above 1 dpa is approximately unity, (2) the stress normalized creep rate is ~1 × 10 –7 [dpa –1 MPa –1 ] at 430–750 °C for the range of 1–30 dpa for most polycrystalline SiC materials, and (3) the effects on irradiation creep of initial microstructures—such as grain boundary, crystal orientation, and secondary phases—increase with increasing irradiation temperature.

36 MATERIALS SCIENCE↗
DOE OSTI2016

Hydrothermal corrosion of silicon carbide joints without radiation

In this paper, hydrothermal corrosion of four types of the silicon carbide (SiC) to SiC plate joints were investigated under pressurized water reactor and boiling water reactor relevant chemical conditions without irradiation. The joints were formed by metal diffusion bonding using molybdenum or titanium interlayer, reaction sintering using Ti—Si—C system, and SiC nanopowder sintering. Most of the joints withstood the corrosion tests for five weeks. The recession of the SiC substrates was limited. Based on the recession of the bonding layers, it was concluded that all the joints except for the molybdenum diffusion bond are promising under the reducing environments without radiation. Finally, the SiC nanopowder sintered joint was the most corrosion tolerant under the oxidizing environment among the four joints.

36 MATERIALS SCIENCE↗
DOE OSTI2017

Irradiation resistance of silicon carbide joint at light water reactor–relevant temperature

We fabricated and irradiated monolithic silicon carbide (SiC) to SiC plate joints with neutrons at 270–310 °C to 8.7 dpa for SiC. The joining methods included solid state diffusion bonding using titanium and molybdenum interlayers, SiC nanopowder sintering, reaction sintering with a Ti-Si-C system, and hybrid processing of polymer pyrolysis and chemical vapor infiltration (CVI). All the irradiated joints exhibited apparent shear strength of more than 84 MPa on average. Significant irradiation-induced cracking was found in the bonding layers of the Ti and Mo diffusion bonds and Ti-Si-C reaction sintered bond. Furthermore, the SiC-based bonding layers of the SiC nanopowder sintered and hybrid polymer pyrolysis and CVI joints all showed stable microstructure following the irradiation.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Stored energy release in neutron irradiated silicon carbide

The purpose of this investigation is to experimentally quantify the stored energy release upon thermal annealing of previously irradiated high-purity silicon carbide (SiC.) Samples of highly-faulted polycrystalline CVD β-SiC and single crystal 6HSiC were irradiated in a mixed spectrum fission reactor near 60 °C in a fluence range from 5 × 10 23 to 2 × 10 26 n/m 2 (E > 0.1 MeV), or about 0.05–20 dpa, in order to quantify the stored energy release and correlate the release to the observed microscopic swelling, lattice dilation, and microstructure as observed through TEM. Within the fluence of this study the crystalline material was observed to swell to a remarkable extent, achieving 8.13% dilation, and then cross a threshold dose for amorphization at approximately 1 × 10 25 n/m 2 (E > 0.1 MeV) Once amorphized the material attains an as-amorphized swelling of 11.7% at this irradiation condition. Coincident with the extraordinary swelling obtained for the crystalline SiC, an equally impressive stored energy release of greater than 2500 J/g at the critical threshold for amorphization is inferred. As expected, following amorphization the stored energy in the structure diminishes, measured to be approximately 590 J/g. Generally, the findings of stored energy are consistent with existing theory, though the amount of stored energy given the large observed crystalline strain is remarkable. The overall conclusion of this work finds comparable stored energy in SiC to that of nuclear graphite, and similar to graphite, a stored energy release in excess of its specific heat in some irradiation conditions.

36 MATERIALS SCIENCE↗