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At least 91 records · Page 5
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Reactive Carbide‐Based Synthesis and Microstructure of NASICON Sodium Metal All Solid‐State Electrolyte

Reactive carbide precursor-based synthesis of NASICON-type NZSP (Na 1+x Zr 2 Si x P 3-x O 12 ) solid-state electrolyte (SSE) is demonstrated, in contrast to the established oxide-based approach. Exothermic decomposition of ZrC and SiC in air homogenizes microstructure, yielding 98% compact density after conventional sintering at 1200 °C. Quantitative stereology demonstrates that significant microstructural differences are present. Compacts of carbide-derived Carb-NZSP are 98% dense with a secondary zirconium oxide (ZrO 2 ) volume fraction of 0.2% ± 0.3%, versus 93% dense and 3% ± 1% for oxide-derived baseline. For Carb-NZSP, the secondary glassy phosphate phase is agglomerated, while for baseline, it is dispersed and percolated. Electrochemical testing combined with post-mortem analysis demonstrates how microstructural control of secondary phases is critical for dendrite suppression: Carb-NZSP critical current density (CCD) is 3.1 ± 0.8 mA cm −2 at 0.1 mAh cm −2 , versus 1.0 ± 0.7 mA cm −2 at 0.1 mAh cm −2 . Cryogenic focused ion beam (cryo-FIB) analysis demonstrates that in both materials, the porous 2D sheet-like sodium metal dendrites propagate around and subsume NZSP grains, likely following a path enriched with glassy phase and with porosity. Dendrites also flow around isolated zirconia particles. Phase field simulation reveals deflection of dendrites by mechanically tough zirconia, while brittle glassy phase accelerates dendrite growth, especially when finely distributed.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Recrystallization suppression through dispersion-strengthening of tungsten

Tungsten is the material of choice for the divertor region of future nuclear fusion reactors, an environment that will expose plasma-facing components (e.g. divertor, etc...) to high temperatures and transient high heat flux events. Under these conditions, recrystallization and grain growth of tungsten can occur, leading to undesirable microstructural and mechanical property changes. Therefore, there is a need to raise the recrystallization temperature of tungsten and limit the kinetics of the recrystallization and grain growth processes. In this work, we examine the effect of different types (TiC vs. TaC vs. ZrC) and different concentrations (1.1 vs. 5 vs. 10 wt.%) of dispersed second phase particles in a tungsten matrix on the high temperature performance. Additionally, the addition of second-phase particles effectively increases the temperature of and time for recrystallization and slow grain growth; however, the addition of a high weight fraction of particles alters the surface chemistry, which may impact subsequent plasma-surface interactions. These results show that the addition of small concentrations of dispersed particles can be effectively employed in tungsten to raise the upper operating temperature limit for tungsten in a fusion reactor.

36 MATERIALS SCIENCE↗
DOE OSTI2025

In-situ irradiation of uranium carbide

Uranium carbide (UC) is a leading candidate fuel for Generation IV reactors due to its high uranium density and thermal conductivity. However, its irradiation performance—particularly gas bubble swelling and defect dynamics—remains poorly characterized. Using in-situ transmission electron microscopy (TEM), we irradiated UC with 300 keV Xe + and 1 MeV Kr 2+ ions at temperatures up to 900 °C to quantify swelling behavior and dislocation loop evolution. The swelling remained below 0.6 % across all temperatures, suggesting the dimensional stability of UC under irradiation at these temperatures. Dislocation loops grew faster in UC than in UO 2 or UN, correlating with its lower homologous temperature. Notably, nanograin structures emerged in thin regions of the lamellar, mirroring phenomena previously observed in UO 2 and ZrC. These results address critical knowledge gaps in the radiation tolerance of UC and provide insight into its suitability for advanced reactor systems.

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 OSTI2026

Overview of advanced plasma-facing materials testing for Fusion Pilot Plants at DIII-D

Characterization and testing of advanced plasma-facing materials (PFMs) for Fusion Pilot Plants (FPP) is being conducted at the DIII-D National Fusion Facility through the ongoing two-year FPP Candidate Materials Thrust. Year one tested 17 novel materials utilizing the Divertor Materials Evaluation System (DiMES), with samples analyzed pre- and post-experiment via SEM, EDS, and confocal microscopy. Repeatable reference discharges were developed to ensure uniformity between experiments, including a new strike-point rastering scenario to provide more uniform heat/particle flux across DiMES during ELMing H-mode discharges. Various sample geometries and temperatures were used to achieve FPP-relevant conditions, including samples angled 10° towards the incident plasma flux and pre-heating up to 500 °C. The first exposure of liquid lithium (Li) capillary porous structures in a tokamak demonstrated uniform emission of Li vapor and suppression of Li droplets in H-mode when preheated to 350 °C. Dispersoid-strengthened W with 1 wt% TaC, TiC, and ZrC exposed to H-mode showed cracking and dispersoid ejection for all varieties except TiC, providing a clear down-selection. Ultra-high temperature ceramic materials TiB 2 and ZrB 2 showed minimal degradation under L-mode exposure. Silicon carbide (SiC) fiber composites showed arcing along edges, while CVD SiC remained pristine. Atmospheric plasma-sprayed W and SiC coatings endured H-mode exposure without macroscopic delamination; SiC exhibited granular ejection, while W showed increased outgassing. Additional W-based alloys were stress tested in H-mode, including Ni-based W heavy alloys, W f SiC f /W composites, W multi-principle element alloys, and functionally-graded W/SiC, to varying degrees of success.

DIII-D↗
DOE OSTI2022

Phase controlled synthesis of transition metal carbide nanocrystals by ultrafast flash Joule heating

Nanoscale carbides enhance ultra-strong ceramics and show activity as high-performance catalysts. Traditional lengthy carburization methods for carbide syntheses usually result in coked surface, large particle size, and uncontrolled phase. Here, a flash Joule heating process is developed for ultrafast synthesis of carbide nanocrystals within 1 s. Various interstitial transition metal carbides (TiC, ZrC, HfC, VC, NbC, TaC, Cr 2 C 3 , MoC, and W 2 C) and covalent carbides (B 4 C and SiC) are produced using low-cost precursors. By controlling pulse voltages, phase-pure molybdenum carbides including β-Mo 2 C and metastable α-MoC 1-x and η-MoC 1-x are selectively synthesized, demonstrating the excellent phase engineering ability of the flash Joule heating by broadly tunable energy input that can exceed 3000 K coupled with kinetically controlled ultrafast cooling (>10 4 K s –1 ). Theoretical calculation reveals carbon vacancies as the driving factor for topotactic transition of carbide phases. The phase-dependent hydrogen evolution capability of molybdenum carbides is investigated with β-Mo 2 C showing the best performance.

01 COAL, LIGNITE, AND PEAT↗
DOE OSTI2023

Effects of transition metal carbide dispersoids on helium bubble formation in dispersion-strengthened tungsten

Abstract The formation of helium bubbles and subsequent property degradation poses a significant challenge to tungsten as a plasma-facing material in future long-pulse plasma-burning fusion reactors. In this study, we investigated helium bubble formation in dispersion-strengthened tungsten doped with transition metal carbides, including TaC, ZrC, and TiC. Of the three dispersoids, TaC exhibited the highest resistance to helium bubble formation, possibly due to the low vacancy mobility in the Group VB metal carbide and oxide phases. Under identical irradiation conditions, large helium bubbles formed at grain boundaries in tungsten, while no bubbles were observed at the interfaces between the carbide dispersoid and tungsten matrix. Moreover, our results showed the interfaces could suppress helium bubble formation in the nearby tungsten matrix, suggesting that the interfaces are more effective in trapping helium as tiny clusters. Our research provided new insights into optimizing the microstructure of dispersion-strengthened tungsten alloys to enhance their performance.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Exploration of LIBS as a novel and rapid elemental mapping technique of nuclear fuels in the form of surrogate TRISO particles

Laser-induced breakdown spectroscopy (LIBS) was employed to characterize coatings on surrogate fuel particles. Tri-structural isotropic (TRISO) particles are a proposed nuclear fuel alternative for high temperature reactors. These particles are constructed of a ZrO 2 kernel (as a surrogate to uranium), surrounded by an inner pyrolytic carbon layer and are surrounded by an outer carbide layer (ZrC, presented here) to act as a barrier to fission products generated during nuclear reactions. These particles are embedded within a graphite compact and housed within the reactor core. Simply put, due to their robust nature, performing elemental analysis of these particles poses a challenge. Presented here, LIBS is explored as a method for characterizing elemental constituents of these particles, with the focus being on rapid elemental mapping and depth profiling. Different from traditional elemental analysis techniques (e.g., inductively coupled plasma – based methods), LIBS is advantageous because it can directly analyze the sample surface and can detect light elements such as C and O, making it a viable technique for the analysis of small, multilayered particles as spatial elemental information is warranted in the production of these particles. In the work presented here, LIBS was successfully used for discerning small layers (30–50 μm), detecting the location of carbon and oxygen layers, providing fast 2-D mapping (<5 min per particle) and rapid depth profiling (10 s per particle).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Pitfalls in parameters: practical process development in chemical vapor processing of SiC

Vapor processing of high-value materials, such as silicon carbide (SiC), is of interest for many industries, including aerospace and energy production. Chemical vapor infiltration (CVI) of additively manufactured components is an especially attractive manufacturing process currently in development. Here, a novel development workflow is demonstrated with the H 2 –CH 3 SiCl 3 gas system for SiC to accelerate the process optimization of CVI SiC. A combination of calculated thermodynamics and high-throughput experimental chemical vapor deposition (CVD) coatings substantially reduced the experiments required with slow CVI processes. The computational results accurately predicted changes in the thermodynamic conditions tested, while CVD coatings – characterized by Raman spectroscopy – addressed changes in kinetic parameters. This workflow is also applicable to other vapor-processing systems, such as pyrolytic carbon, ZrC, or Si 3 N 4 .

Lamm, Benjamin W. [Oak Ridge National Laboratory (↗
DOE OSTI2024

Compositionally complex carbide ceramics: A perspective on irradiation damage

Extensive experimental and computational studies have demonstrated outstanding physical and chemical properties of the novel materials of compositionally complex carbides (CCCs), enabling their promising applications in advanced fission and fusion energy systems. This perspective provides a comprehensive overview of radiation damage behavior reported in the literature to understand the fundamental mechanisms related to the impact of multi-principal metal components on phase stability, irradiation-induced defect clusters, irradiation hardening, and thermal conductivity of compositionally complex carbides. Several future research directions are recommended to critically evaluate the feasibility of designing and developing new ceramic materials for extreme environments using the transformative “multi-principal component” concept. Compared to the existing materials for nuclear applications including stainless steels, nickel alloys, ZrC, SiC, and potentially high-entropy alloys, as well as certain other compositionally complex ceramic families. CCCs appear to be more resistant to amorphization, growth of irradiation defect clusters, and void swelling.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Chemical Vapor Deposition of Zirconium Compounds: A Review

Coatings of zirconium compounds are used in a wide variety of fields, yet an understanding and descriptions of deposition mechanisms are scant in the public literature. The mechanisms of deposition for metallic zirconium, ZrC, ZrN, ZrO 2 , ZrB 2 , and zirconium silicides are discussed based on the direct vapor deposition research of those compounds where possible or compared to complementary titanium systems when direct research is lacking. Both inorganic and organometallic deposition systems are discussed. As a class of compounds, an understanding of the vapor deposition mechanisms can be significantly improved by investigations on metallic zirconium deposition by zirconium halides and hydrogen and by in situ analysis techniques such as Fourier-transform infrared (FTIR) spectroscopy or x-ray photoelectron spectroscopy (XPS).

organometallic↗
DOE OSTI2023

Hot Isostatic Pressing Control of Tungsten-Based Composites

Metal-oxide composites are commonly used in high temperature environments for their thermal stability and high melting points. Commonly employed with refractory oxides or carbides such as ZrC and HfC, these materials may be improved with the use of a low density, high melting point ceramic such as CeO2. In this work, the consolidation of W-CeO2 metal matrix composites in the high CeO2 concentration regime is explored. The CeO2 concentrations of 50, 33, and 25 wt.%, the CeO2 particle size from nanometer to micrometer, and various hot isostatic pressing temperatures are investigated. Decreasing the CeO2 concentration is observed to increase the composite density and increase the Vickers hardness. The CeO2 oxidation state is observed to be a combination of Ce3+ and Ce4+, which is hypothesized to contribute to the porosity of the composites. The hardness of the metal-oxide composite can be improved more than 2.5 times compared to pure W processed by the same route. This work offers processing guidelines for further consolation of oxide-doped W composites.

36 MATERIALS SCIENCE↗
NASA NTRS1972

Fabrication of (U, Zr) C-fueled/tungsten-clad specimens for irradiation in the Plum Brook Reactor Facility

Fuel samples, 90UC - 10 ZrC, and chemically vapor deposited tungsten fuel cups were fabricated for the study of the long term dimensional stability and compatibility of the carbide-tungsten fuel-cladding systems under irradiation. These fuel samples and fuel cups were assembled into the fuel pins of two capsules, designated as V-2E and V-2F, for irradiation in NASA Plum Brook Reactor Facility at a fission power density of 172 watts/c.c. and a miximum cladding temperature of 1823 K. Fabrication methods and characteristics of the fuel samples and fuel cups prepared are described.

Source record↗
NASA NTRS1973

Fabrication and life testing of thermionic converters

An unfueled converter containing a chloride-fluoride duplex tungsten emitter of 4.78 eV vacuum work function was tested for 46,647 hours at an emitter temperature of 1973 K and an electrode power output of about 8 watts/sq cm. The test demonstrated the superior and stable performance of the (110) oriented tungsten emitter at high temperatures. Three 90 UC-10 ZrC(C/U = 1.04, tungsten additive = 4 wt %) fueled converters were fabricated and tested at an emitter temperature of 1873 K. Converter containing chloride-arc-cast duplex tungsten cladding showed temperature thermionic performance and slower rate of performance drop than converter containing chloride-fluoride duplex tungsten cladding. This is believed to be due to the superior fuel component diffusion resistance of the arc-cast tungsten substrate used in the fuel cladding. It was shown that a converter containing a carbide fueled chloride-arc-cast duplex tungsten emitter with an initial electrode power output of 6.80 watts/sq cm could still deliver an electrode power output of 6.16 watts/sq cm after 18,632 hours of operation at an emitter temperature of 1873 K.

Yang, L.↗
NASA NTRS1973

NERVA irradiation program. GTR 23, volume 1: Combined effects of reactor radiation and cryogenic temperature on NERVA structural materials

Specimens fabricated from structural materials that were candidates for certain NERVA applications were irradiated in liquid nitrogen (LN2), liquid hydrogen (LH2), water, and air. The specimens irradiated in LN2 were stored in LN2 and finally tested in LN2, or at some higher temperature in a few instances. The specimens irradiated in LH2 underwent an unplanned warmup while in storage so this portion of the test was lost; some specimens were tested in LN2 but none were tested in LH2. The Ground Test Reactor was the radiation source. The test specimens consisted mainly of tensile and fracture toughness specimens of several different materials, but other types of specimens such as tear, flexure, springs, and lubricant were also irradiated. Materials tested include Hastelloy X, Al, Ni steel, steel, Be, ZrC, Ti-6Al-4V, CuB, and Ti-5Al-2.5Sn.

Mcdaniel, R. H.↗
NASA NTRS1974

Evaluation of cermet materials suitable for lithium lubricated thrust bearings for high temperature operation

Cerment materials (HfC - 10 wt% W; HfC - 10 wt% TaC - 10 wt%W; HfC - 2 wt% CbC - 8 wt% Mo;Hfn - 10 wt% W; Hfn - 10 wt% TaN - 10 wt% W; and ZrC - 17 wt% W) were evaluated for possible use as lithium-lubricated bearings in the control system of a nuclear reactor. Tests of compatibility with lithium were made in T-111 (Ta-8W-2Hf) capsules at temperatures up to 1090 C. The tendencies of HfC-TaC-W, HfC-CbC-Mo, and HfN-W to bond to themselves and to the refractory alloys T-111 and TZM when enclosed in lithium-filled capsules under a pressure of 2000 psi at 980 and 1200 C for 1933 hours were evaluated. Thermal expansion characteristics were determined for the same three materials from room temperature to 1200 C. On the basis of these tests, HfC-10 TaC-10W and HfN-10W were selected as the best and second best candidates, respectively, of the materials tested for the bearing application.

Sinclair, J. H.↗
NASA NTRS1975

Carbide coated fibers in graphite-aluminum composites

The NASA-supported program at the Los Alamos Scientific Laboratory (LASL) to develop carbon fiber-aluminum matrix composites is described. Chemical vapor deposition (CVD) was used to uniformly deposit thin, smooth, continuous coats of TiC on the fibers of graphite tows. Wet chemical coating of fibers, followed by high-temperature treatment, was also used, but showed little promise as an alternative coating method. Strength measurements on CVD coated fiber tows showed that thin carbide coats can add to fiber strength. The ability of aluminum alloys to wet TiC was successfully demonstrated using TiC-coated graphite surfaces. Pressure-infiltration of TiC- and ZrC-coated fiber tows with aluminum alloys was only partially successful. Experiments were performed to evaluate the effectiveness of carbide coats on carbon as barriers to prevent reaction between alluminum alloys and carbon. Initial results indicate that composites of aluminum and carbide-coated graphite are stable for long periods of time at temperatures near the alloy solidus.

Imprescia, R. J.↗