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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 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Mechanical properties of fusion welded ceramics in the $\mathrm{SiC-ZrB}$ 2 and $\mathrm{SiC-ZrB}$ 2 -$\mathrm{ZrC}$ systems

Mechanical properties of welded SiC-ZrB 2 and SiC-ZrB 2 -ZrC ceramics were measured up to 1700 °C. Commercial powders were hot pressed, machined into coupons, and preheated to 1600 °C before joining the ceramics using either tungsten inert gas welding or plasma arc welding. Toughness of the parent materials was 3–4 MPa*m 1/2 which decreased after welding to 2–2.5 MPa*m 1/2 . Strength of the SiC-ZrB 2 -ZrC parent material was ~700 MPa at 25 °C, ~300 MPa at 1700 °C, and retained 40–60% of this strength once welded. Strength of the SiC-ZrB2 parent material was ~600 MPa at 25 °C and 1700 °C and retained 20–30% of this strength once welded. Griffith analysis indicated that the strength in the parent materials was controlled by the size of SiC clusters while strength of welds was controlled by the size of pores in fusion zones. Therefore, removal of pores in produced fusion zones should be investigated to improve strength of future ceramic welds.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Response of 11 B enriched ZrB 2 ultra-high temperature ceramic to neutron irradiation at elevated temperatures

ZrB 2 , an ultra-high temperature ceramic (UHTC) is being considered for use in fusion reactor first-wall structures, yet its response to irradiation remains poorly understood. This study employed scanning/transmission electron microscopy (S/TEM), synchrotron X-ray diffraction (XRD), finite element calculations, and thermal property measurements to thoroughly investigate the neutron-irradiation effects on 11 B-enriched ZrB 2 . Neutron irradiations were conducted at 220 °C and 620 °C, with a neutron fluence of 2.2 × 10 25 neutron/m 2 (energy > 0.1 MeV), resulting in 3.9 dpa and 4200 appm He. The study revealed the unusual prevalence of prism loops and a > c anisotropic lattice swelling, likely linked to the low c/a ratio of ZrB 2 , leading to grain boundary microcracking. Reducing the grain sizes was effective in reducing intergranular cracking and macroscopic swelling. The observation of cavities in ZrB 2 irradiated at 620 °C, as opposed to 220 °C, prompts questions about the temperature at which vacancies in ZrB 2 become mobile, and the role of neutron absorption by 10 B in elevating irradiation temperatures. Isotopic enrichment in 11 B proves to be a viable strategy for mitigating helium production in transition-metal diborides, which is a critical consideration for nuclear applications. Irradiation-induced defects reduce the thermal diffusivity and conductivity of ZrB 2 by a factor of 4–9, which has important implications for its role as a plasma-facing material in fusion reactors that drive high heat fluxes through first-wall materials. Here, this comprehensive study lays the foundation for understanding ZrB 2 behavior under neutron irradiation and highlights important phenomena to consider for various material applications.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Characterization of fusion welded ceramics in the SiC-ZrB 2 -ZrC system

Various SiC-ZrB 2 -ZrC ceramics were joined by fusion welding to determine the maximum silicon carbide content that could be joined. Commercial powders were hot pressed, machined, and preheated to 1450 °C before joining with a tungsten inert gas welding torch at 160–200 A. Resulting welds were cross-sectioned and analyzed to determine which compositions were weldable and to characterize microstructural evolution in welded samples. As compositions approached the ternary eutectic, the welds had smaller SiC grains and exhibited better weldability. Penetration depth of welds was controlled by a combination of current input and welding speed. Here, the ternary eutectic in the system was found at 36.9 ± 1.3 vol% SiC, 42.7 ± 1.5 vol% ZrB 2 , and 20.4 ± 1.9 vol% ZrC and its melting temperature was 2330 ± 23 °C. A ternary phase diagram for the SiC-ZrB 2 -ZrC was constructed and proposed via microstructural analysis of arc melted pellets on binary joins between each binary eutectic and the ternary eutectic in the system.

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 OSTI2024

Deuterium plasma induced preferential erosion in ultra-high temperature ceramics TiB 2 and ZrB 2 *

Abstract Steady-state deuterium plasma exposures were performed on ultra-high temperature ceramics titanium diboride (TiB 2 ) and zirconium diboride (ZrB 2 ) using the PISCES-RF linear plasma device (LPD) as early screening for first wall, plasma-facing applications. Deuterium plasma exposures were performed using 40 eV ion energies at 240, 525, and 800 °C sample temperatures and 90 eV ion energies at 240 °C sample temperatures to analyze TiB 2 and ZrB 2 surface morphology and chemistry evolution behavior. Post-plasma exposure chemistry characterization of the near surface ( < 50 nm) region of the samples all show transition metal enrichment, indicating boron preferential erosion. Transition metal to boron fractions vary with plasma exposure temperature under the 40 eV ion energy; metal enrichment is maximized at 800 °C and then minimized at 525 °C. SEM micrographs of all plasma exposed sample surfaces show no significant or noticeable plasma induced damage from cracking or blistering.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2020

Materials Data on ZrB by Materials Project

ZrB is Halite, Rock Salt structured and crystallizes in the cubic Fm-3m space group. The structure is three-dimensional. Zr3+ is bonded to six equivalent B3- atoms to form a mixture of edge and corner-sharing ZrB6 octahedra. The corner-sharing octahedral tilt angles are 0°. All Zr–B bond lengths are 2.46 Å. B3- is bonded to six equivalent Zr3+ atoms to form a mixture of edge and corner-sharing BZr6 octahedra. The corner-sharing octahedral tilt angles are 0°.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Materials Data on ZrB by Materials Project

ZrB is Caswellsilverite-derived structured and crystallizes in the hexagonal P-6m2 space group. The structure is three-dimensional. there are two inequivalent Zr3+ sites. In the first Zr3+ site, Zr3+ is bonded in a 6-coordinate geometry to six equivalent B3- atoms. All Zr–B bond lengths are 2.42 Å. In the second Zr3+ site, Zr3+ is bonded in a 6-coordinate geometry to six equivalent B3- atoms. All Zr–B bond lengths are 2.51 Å. B3- is bonded to six Zr3+ atoms to form a mixture of corner, edge, and face-sharing BZr6 octahedra. The corner-sharing octahedra tilt angles range from 49–51°.

36 MATERIALS SCIENCE↗
DOE OSTI2026

Comparative neutron-irradiation effects on thermal conductivity degradation and dimensional stability of TiC, TiB 2 , and ZrB 2 at 200–1000 °C

Ultra-high-temperature ceramics (UHTCs), including TiC, Ti 11 B₂, and Zr 11 B₂, show great potential for plasma-facing components due to their excellent high-temperature properties prior to irradiation. However, their response to neutron irradiation remains insufficiently understood, limiting robust assessment of their viability for fusion energy applications. Here, this study examines the thermal conductivity, dimensional stability and microstructure of TiC, TiB₂, and ZrB₂ following neutron irradiation at temperatures of 200–1000 °C and fast neutron fluences of 2.0 × 10 25 to 1.1 × 10 26 n/m 2 (E > 0.1 MeV). Lattice swelling measured by synchrotron X-ray diffraction in all three UHTCs was maximized at 200 °C and decreased with increasing irradiation temperature, with no evidence of amorphization observed at 200 °C. Above 600 °C, significant macroscopic volume swelling was observed in irradiated Ti 11 B₂ and Zr 11 B₂, but not in TiC, likely due to cavity formation in the diborides. The post-irradiation thermal conductivity, measured at the irradiation temperature, ranged from 28 to 45 W/m·K, representing a 34–45% reduction relative to the unirradiated material. Notably, neutron-irradiated UHTCs exhibit recoverable thermal conductivity at elevated temperatures, comparable to ferritic–martensitic steels and potentially superior to W when transmutation effects are considered, highlighting promise for shielding or armor plasma-facing components. At 600 °C, both thermal conductivity degradation and lattice swelling saturated at doses exceeding 2–4 dpa.

fusion materials↗
DOE OSTI2022

Fusion welding of refractory metals and ZrB 2 -SiC-ZrC ceramics

Molybdenum and a molybdenum alloy were fusion welded to ZrB2-based ceramics to determine if the electrical and thermal properties of the metals and ceramics affected their weldability. Commercial ceramic powders were hot pressed, machined into coupons, and preheated to 1600 °C before joining the ceramics to commercial metals using plasma arc welding. Weldability varied as indicated by the range of porosity observed within the fusion zones. Measured thermal and electrical properties appeared to have little to no effect on the weldability of metal-ceramic welds despite the large range of values measured across each property. Differences in melting temperatures between metal and ceramic coupons did affect weldability by changing the weld penetration depth into ceramic coupons. Finally, future studies on metal-ceramic welds are suggested to investigate the effect that work function, melt viscosity, wetting, or other properties have on weldability.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Zirconium-diboride silicon-carbide composites: A review

Zirconium diboride (ZrB 2 ) and silicon carbide (SiC) composites have long been of interest since it was observed that ZrB 2 improved the thermal shock resistance of SiC. However, processing of these materials can be difficult due to high and different sintering temperatures and differences in the thermodynamic stability of each material. ZrB 2 –SiC composites have been processed in a variety of ways including hot-pressing, spark-plasma sintering, reactive melt infiltration, pack cementation, chemical vapor deposition, chemical vapor infiltration, stereolithography, direct ink writing, selective laser sintering, electron beam melting, and binder jet additive manufacturing. Each manufacturing method has its own pros and cons. This review serves to summarize more than 60 years of research and provide a coherent resource for the variety of methods and advancements in development of ZrB 2 –SiC composites.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Structural Tolerance of Zirconium Diboride under Electron Irradiation through in-situ Convergent Beam Electron Diffraction and Energy-dispersive X-ray Spectroscopy

Zirconium diboride (ZrB 2 ) is a ultra-high temperature ceramic with high melting temperature (>3000 K), super strength and hardness, good thermal conductivity, and excellent resistance to plasma erosion and chemical erosion, being widely applied in hypersonic flight, arc-electrodes, refractory materials, and plasma-facing materials in fusion reactors. Therefore, the structural stability of the material is essential for the applications. Up to now, the material was investigated under neutron irradiation and high-energy ion irradiations. It was reported that the neutron irradiation caused swelling and catastrophic cracking. High-energy ions (30 keV He + ) changed the surface morphology and caused the bright blue and brown colors on ZrB 2 surfaces between 920 K and 1120 K. The electron irradiation tolerance of the material has not been reported yet. Here we used convergent beam electron diffraction (CBED) and energy-dispersive X-ray spectroscopy (EDS) to investigate structural tolerances of ZrB 2 crystals under high-energy electrons at room temperature under vacuum. Finally, the investigation would benefit the understanding of its structural tolerance at microscales.

42 ENGINEERING↗
DOE OSTI2023

Size effects in 3D–printed polymer–derived, zirconium diboride–reinforced ceramic composites

Preceramic polymers are of interest for use in many manufacturing techniques such as injection molding, ceramic fiber infiltration, and additive manufacturing. However, off-gassing of low molecular weight oligomers occurs when these polymers cure, potentially leading to porosity in the cured part. Here to study how porosity and strength are affected by the size of the printed part, and the presence of a high surface area nano-scale filler, polycarbosilane (PCS) microrods of varying diameter were fabricated via direct ink writing (DIW), an additive manufacturing technique, with two ink formulations containing either zirconium diboride (ZrB 2 ) alone, or ZrB 2 and fumed alumina (FA). Sets of microrods were printed in a range of sizes by using print nozzles of 450, 634, 979, 1 346, or 1 702 μm in diameter, which were thermally cured, pyrolyzed to form ceramic composite microrods, and tested in 3-pt flexure. Porosity increased with increasing diameter, while failure strength decreased. For a given nozzle size, the microrods containing FA displayed lower porosity and higher strength (up to ~500 MPa) compared to the microrods containing only ZrB 2 . Weibull strength analysis was performed on each group of microrods and shows that the addition of FA increased Weibull modulus from 4.63 ± 1.56 to 9.35 ± 0.601. In conjunction with optical microscopy, this analysis indicates two distinct flaw populations in the printed materials, porosity which arises during the curing step and cracking which arises during pyrolysis of the larger specimens.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Additive Manufacturing of Corrosion Resistant UHTC Materials for Chloride Salt-to-sCO2 Brayton Cycle Heat Exchangers

The objectives of the project include: (1) screen high-purity, high density ultra-high-temperature ceramics (UHTCs) for chloride salt corrosion resistance, (2) develop binder-jet additive manufacturing and sintering for ZrB 2 -MoSi 2 blends using literature and experiments to demonstrate feasibility, and (3) evaluate chloride salt corrosion resistance of additively manufactured ZrB 2 -MoSi 2 . These initial objectives support the development of a durable and reliable ultra-compact high-efficiency salt-to-sCO 2 heat exchanger for closed-loop Brayton power cycles based on corrosion resistant, high-strength, thermally conductive UHTC materials and novel designs enabled by additive manufacturing. The UHTC materials are also abrasion resistant so that they could be applicable to all three thermal carrier media pathways identified in the CSP Gen3 Roadmap.

36 MATERIALS SCIENCE↗
DOE OSTI2026

Water Vapor Resistant SiC/SiC Composite for Hydrogen-Based Turbines (Final Scientific Technical Report)

New material innovations are needed for the extreme environments of next generation, hydrogen-fueled turbine engines. Ceramic matrix composites (CMCs) offer high-temperature capability but are susceptible to degradation in high-temperature water vapor. Pratt & Whitney’s (P&W) baseline silicon carbide (SiC)/SiC CMC was initially applied to design a component in a hydrogen-fueled engine, verifying the existence of a design space and extracting boundary conditions to inform testing parameters. Several material innovations, including two fibers, interface coatings (IFC), and self-healing matrices (SHM), were investigated to improve high-temperature performance in water vapor. A boron-doped pyrocarbon (B-PyC) IFC and a SHM matrix with layers of zirconium nitride (ZrN) or zirconium diboride (ZrB 2 ) were developed to fabricate minicomposites with either standard Hi-Nicalon™ Type S (HNS) fibers or new Tyranno® SA4 fibers (SA4). The B-PyC IFC is functional but lacks in providing improved durability. A matrix consisting of thick layers of SiC and thin layers of ZrB 2 shows promise as a SHM that can effectively seal matrix cracks in this extreme environment. Minicomposites with HNS fibers generally outperform those with SA4 fibers.

08 HYDROGEN↗
DOE OSTI2022

Binder jet additive manufacturing of ceramic heat exchangers for concentrating solar power applications with thermal energy storage in molten chlorides

Triply periodic minimal surface (TPMS) geometries can only be fabricated by additive manufacturing methods and are of interest for heat exchangers. Ceramic TPMS heat exchangers can operate at higher temperatures and pressures with superior performance and increased operating efficiencies compared to metal heat exchangers. The properties of ultra-high temperature ceramic (UHTC) materials are also favorable for heat exchangers and potentially suited to concentrated solar power (CSP) systems, such as those based on a molten chloride salt thermal energy storage (TES) medium used to heat CO 2 in a closed-loop Brayton power cycle. We intended to demonstrate binder jet additive manufacturing feasibility of a UHTC-TPMS structure by printing and sintering a null candidate. We aimed to achieve parts with a relative density ≥ 92 % of theoretical and to provide a TPMS part demonstration. The target density indicates the transition from intermediate to final stage sintering, a requirement to inhibit gas permeability and for sintering complex near net shapes to full density with sinter-HIP technology. The goal of the TPMS part demonstration was to determine if printing and sintering parameters developed from test coupons apply to the complex geometries that will eventually be used in a heat exchanger design. Our objective was to print cubic TPMS parts with a 9 cm 3 volume and sinter it without distorting and cracking. We report we were able to sinter ZrB 2 -MoSi 2 composite parts based on the Schwarz-D TPMS and achieve isotropic shrinkage up to 60 % by volume, resulting in densities ranging from 92 % to 96 % of theoretical.

36 MATERIALS SCIENCE↗
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 OSTI2023

Reactivity-initiated accidents in two pressurized water reactor high burnup core designs

Here, this paper presents a safety analysis of two proposed core loadings for 24-month Pressurized Water Reactor (PWR) fuel cycles. This analysis focuses on reactivity-initiated accidents (RIAs) and evaluates core safety performance impacts of rod-averaged burnup limits up to 75 GWD/MTU and less than 7 % enriched UO 2 . The capabilities of Polaris, PARCS, and RELAP5-3D are leveraged to evaluate the core neutronic and thermal–hydraulic behavior for normal-operation, uncontrolled control rod withdrawal (CRW) transients, and control rod ejection (CRE) accidents. The two core designs are compared to identify features of realistic high burnup/extended enrichment core design approaches which have significant safety impact, identify experimental data needs for high-fidelity predictive modeling, and provide recommendations for future high burnup core designs. The first core design evaluated in this study was developed by Southern Nuclear Company and used an ZrB 2 Integral Fuel Burnable Absorber (IFBA) and B 4 C Wet-Annular Burnable Absorber (WABA)-based burnable poison strategy. The second core design assessed in this work used a Gd 2 O 3 -doped UO 2 burnable poison, similar to that used in boiling water reactors or French PWRs. Results indicate that fuel thermal limits are maintained for limiting CRW and hot full power (HFP) CRE transients. Cladding failure is predicted for the highest energy deposition rods in each core during limiting hot zero power (HZP) CRE accidents (where maximum radially averaged enthalpy exceeds 120 cal/g), though licensing may be permissible with a limited number of failed rods. While concerns exist regarding high critical boron concentration during steady state for the IFBA core and large plenum pressures for the gadolinia core design, the analysis demonstrates adequate safety performance during limiting RIA accident scenarios for two representative high burnup core designs. Design changes limiting plenum pressures and implementation of accident tolerant fuel (ATF) cladding features which minimize hydriding and susceptibility to pellet-cladding mechanical interaction (PCMI) are recommended for future high burnup fuel concepts. To support the technical basis for burnup limit increases, high-fidelity fuel performance models are needed to address physical effects not considered in this analysis, and high burnup irradiated fuel tests are required to extend applicability of the fuel failure limits and validate existing and future models.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗