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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 55 records · Page 3
All sourcesNASA NTRSDOE OSTI
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 OSTI2024

Alloying effects on the microstructure and properties of laser additively manufactured tungsten materials

A large body of literature within the additive manufacturing (AM) community has focused on successfully creating stable tungsten (W) microstructures due to significant interest in their application for extreme environments. However, cracking and additional embrittling features at grain boundaries have resulted in poorly performing materials, stymying the application of AM as a manufacturing technique for W. Several alloying strategies, such as ceramic particles and ductile elements, have emerged with the promise to eliminate cracking while simultaneously enhancing stability against recrystallization. Here, in this work, we provide new insights regarding the defects and microstructural features that result from the introduction of ZrC for grain refinement and NiFe as a ductile reinforcement phase – in addition to the resulting thermophysical and mechanical properties. ZrC is shown to promote microstructural stability with increased hardness due to the formation of ZrO 2 dispersoids. Conversely, NiFe forms into micron-scale FCC phase regions within a BCC W matrix, producing enhanced toughness relative to pure AM W. A combination of these effects is realized in the WNiFe + ZrC system and demonstrates that complex chemical environments coupled with the tuning of AM microstructures provides an effective pathway for enabling laser AM W materials with enhanced stability and performance.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Thermal Neutron Scattering Law Evaluations for Zirconium Carbide and Critical Mass Calculations [Slides]

Zirconium Carbide (ZrC) is a refractory material currently being investigated for high temperature reactor applications. ZrC was studied as a base of (U,Zr)C solid solution and UC 2 composite fuels for NERVA-type cores during the Rover program. Testing of the material was performed at Nuclear Furnace 1 (NF-1) located at Los Alamos. Thermal Scattering Law (TSL) evaluations generated with FLASSH are a new contribution to ENDF/B-VIII.1.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2021

Characterization of zirconium carbide microspheres synthesized via internal gelation

Microspheres of zirconium carbide with grain size 40-320 nm were synthesized using internal gelation techniques and characterized by scanning electron microscopy, synchrotron X-ray diffraction, and neutron total scattering. Compared with a polycrystalline benchmark ZrC sample prepared by plasma vapor phase deposition, the microspheres display lower variance in grain size and considerably lower microstrain representative of a more homogenous internal microstructure. However, excess carbon was present in the microspheres, evident in both the X-ray and neutron diffraction data as well as a corresponding hypostoichiometric ZrC phase. Furthermore, the excess carbon phase is assumed to be pockets of carbon that remained unreacted through the combined internal gelation and subsequent carbothermic reduction synthesis process. Atomic-scale structural characterization with neutron PDF analysis confirmed the presence of localized nano-sized domains of graphite-like carbon material.

36 MATERIALS SCIENCE↗
DOE OSTI2025

TEM characterization of two variants of fuel cladding chemical interaction in a HT-9 Clad U-10Zr Fuel. Variant 1: FCCI with a Zr Rind

Here, this study investigated the fuel cladding chemical interaction (FCCI), a key factor that limits operational temperature and burnup, in an HT-9 clad U-10Zr nuclear fuel sample irradiated to a high burnup of 13.1 at.% at a time-averaged peak inner cladding temperature (PICT) of 530 °C. Previous results showed this fuel sample exhibited two distinct levels of FCCI at d. This paper analyzed the FCCI at an azimuthal position showing an interdiffusion layer of <10 µm using transmission electron microscopy to examine chemical and crystallographic nature of phases at the fuel-cladding interface at the nanoscale level. A ZrC layer and a Zr 3 Si phase were identified at the interface; these, along with the relatively low local temperature, potentially contributed to limit interdiffusion, behaving as inhibitors for deleterious interactions. Lanthanides (Ln) partially consumed the ZrC layer and interacted with Fe, forming a Zr-Ln compound and a (Zr,Ce)Fe 2+x phase while also infiltrating up to 4 µm into the cladding. Neither U nor Zr were observed in the cladding, whereas Fe diffused up to 3–5 µm in the fuel. Fe infiltration formed a ternary U-Zr-Fe ε-phase and likely promoted the precipitation of a Cr-rich α’ phase on the cladding interface. Additionally, a Cr-rich χ-phase, likely formed by the dissociation of pre-existing M 23 C 6 carbide precipitates, was identified about 2–5 µm from the fuel-cladding interface. Irradiation-induced nano-voids were also observed in the HT-9 bulk. These findings provide critical insights into FCCI mechanisms at representative irradiation conditions, essential for developing models simulating in-pile metallic fuel behaviors for next-generation reactors.

36 - MATERIALS SCIENCE↗
DOE OSTI2021

Processing and microstructure of ZrB 2 –SiC composite prepared by reactive spark plasma sintering

In-situ formation of ZrB 2 –SiC composites was investigated by reactive spark plasma sintering of precursor powders according to the reaction B4C + 2ZrC + 3Si → 3SiC + 2ZrB 2 . The reaction and process presented here involves a diffusion reaction between B4C and ZrC which facilitates the formation of ZrB 2 , while liquid phase sintering of silicon facilitates atomic diffusion and combines with free C from the B4C and ZrC reaction to form SiC within minutes of heating and there were some residual unreacted precursor materials. An interpenetrating matrix of ZrB 2 –SiC was formed that shows increased fracture toughness (6.03 ± 0.45 MPa m 1/2 ) despite relatively low density (95 %).

36 MATERIALS SCIENCE↗
DOE OSTI2021

Carbide Nanoparticle Dispersion Techniques for Metal Powder Metallurgy

Nanoparticles (NP) embedded into a matrix material have been shown to improve mechanical properties such as strength, hardness, and wear-resistance. However, the tendency of NPs to agglomerate in the powder mixing process is a major concern. This study investigates five different mechanochemical processing (MCP) routes to mitigate agglomeration to achieve a uniform dispersion of ZrC NPs in an Fe-based metal matrix composite. Our results suggest that MCP with only process controlling agents is ineffective in avoiding aggregation of these NPs. Instead, the uniformity of the carbide NP dispersion is achieved by pre-dispersing the NPs under ultrasonication using suitable surfactants followed by mechanically mixing of the NPs with iron powders in an alcohol solvent which is then dried. High-energy MCP is then used to embed the NPs within the powders. These collective steps resulted in a uniform dispersion of ZrC in the sintered (consolidated) Fe sample.

08 HYDROGEN↗
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↗