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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 163 records · Page 9
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Multiscale Modeling of the Mechanical Response of Silicon Carbide Composite Within the Accelerated Fuel Qualification Framework

The accelerated fuel qualification (AFQ) framework has been used for the initial development of multiscale modeling of silicon carbide (SiC) fiber reinforced composite (SiC-SiC). The AFQ framework provides a methodology to leverage physics-informed multiscale modeling along with a reduced set of empirical test data to reduce the time and cost of licensing and qualification of new nuclear fuel systems while maintaining the overall nuclear power plant safety case. SiC-SiC is being proposed for in-core applications, most notably fuel cladding, for current and next-generation nuclear reactors because of its high temperature stability, irradiation tolerance, and ability to withstand many accident conditions. As these composites exhibit multiscale architectures and complex microstructure-based fracture mechanics, it is an appealing use case for the AFQ methodology. While the end goal of this work is a single multiscale model that can be used for predictive in-core performance, current focus is on the individual various length scale models. Four individual models have been initially developed from microscale to engineering system level to capture key physics-based effects across different length scales. These models include a microscale homogenized tow model, a mesoscale fast Fourier transform–based weave model that integrates the homogenized tow model, a mesoscale finite element–based weave model, and a system-level BISON fuel performance model. Results of these models have undergone an initial comparison with separate-effects test data showing a good match to experimental results. By using the AFQ framework during model development, several near-term benefits have been secured including a reduction in development time for the SiC-SiC cladding, more targeted irradiation testing, and a better understanding of uncertainty.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Microstructural and infiltration properties of woven preforms during chemical vapor infiltration

Interface-resolved direct numerical simulations (DNSs) of chemical vapor infiltration (CVI) have been performed over a range of furnace-operating conditions (Thiele moduli) and for practical woven preform geometries. A level-set method is used to resolve the geometry of the initial preform at tow scale. The interface between the vapor and solid phase is then evolved in time through the entire CVI densification cycle, fully resolving the time-varying topology between the two phases. In contrast to previous level-set methods for CVI simulation, the physical reaction and diffusion processes govern the level-set movement in the current approach. The surface deposition kinetics is described by the usual one-step model. In this paper, the DNS data are used to study the evolving porosity, surface-to-volume ratio, and flow infiltration properties (permeability and effective diffusivities). Additionally, comparisons are made to popularly-assumed structure functions and the standard, Kozeny–Carmen porous media model commonly employed in modeled CFD simulations of CVI. The virtual DNS experiments reveal a Thiele modulus and preform geometry (fabric layup) dependence which the existing microstructural and infiltration models are not able to describe throughout the entire densification process. The DNS-based, woven geometry-specific correlations can be applied directly to mean-field, furnace-scale CFD simulations.

36 MATERIALS SCIENCE↗
DOE OSTI2026

Efficient Silicon Carbide (SiC) Fiber Manufacture: Continuous Processing of Novel Precursors via Modified Material Handling

General Atomics Electromagnetic Systems (GA-EMS) together with The Nonwovens Institute (NWI) and Harper International have executed on a development project to improve the production efficiency and reduce the production cost of silicon carbide (SiC) fiber – a critical material that advances US energy security and aerospace leadership. In the first budget period (BP1), the modified tow throughput (MTT) apparatus was developed, enabling fiber spinning in the form of loose coils rather than tightly packed spools (Figure 1) for batch processing of SiC fiber. This coil-based approach was proven to be capable of uniform fiber crosslinking and ceramic conversion, as well as demonstrating key process efficiencies while meeting the program fiber property targets.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Ceramic–Metal (Cermet) Composites: A Review of Key Properties and Synthesis Methods Focused on Nuclear Waste Immobilization

Here this paper reviews key properties, applications, and examples of ceramic-metal composites (cermets), and metal matrix composites (MMC) with emphasis on their applicability as waste forms for immobilizing nuclear waste. While the literature is mature for vitrified and cementitious radioactive waste forms, cermet materials have not received adequate attention as potential candidates for immobilizing nuclear waste. To promulgate this effort, this review connects cermet and MMC design, such as hardened tools, with the chemistry of radioactive waste streams. A discussion on certification and qualification standards for cermet waste forms, literature gaps, and “how-to” sections on cermet processing techniques. Key parameters discussed include thermal conductivity, chemical durability, and waste loading, as well as examples for metal-containing waste forms. As cermet waste forms gain momentum within the community, this review paper aims ensure the end-of-life nuclear fuel cycle is addressed from a materials and waste disposition perspective.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2021

Immobilization of cesium and iodine into Cs 3 Bi 2 I 9 perovskite-silica composites and core-shell waste forms with high waste loadings and chemical durability

Cs$_3$Bi$_2$I$_9$, a defect perovskite derivative, is a potential host phase to immobilize iodine and cesium with high waste loadings. In this work, two strategies were explored to form Cs$_3$Bi$_2$I$_9$-silica composites and a core-shell structure in order to improve chemical durability of waste form materials meanwhile maintaining high waste loadings. Cs$_3$Bi$_2$I$_9$ loadings as high as 70 wt.% were incorporated into a silica matrix to form silica-ceramic composites, and 20 wt.% Cs$_3$Bi$_2$I$_9$ was encapsulated into silica to form a core–shell structure by low temperature spark plasma sintering. Chemical durability of the composite and core-shell waste forms was evaluated by semi-dynamic leaching experiments, and Cs and I were incongruently released from waste form matrices. A BiOI alteration layer formed, acting as a passivation layer to reduce the release of radionuclides. The long-term iodine release rate was low (30 mg m$^{-2}$ day$^{-1}$) for the 70 wt.%Cs$_3$Bi$_2$I$_9$–silica composite leached in deionized water at 90 °C, which can be further reduced to 5 × 10$^{-3}$ mg m$^{-2}$ day$^{-1}$ for the 20 wt.% core-shell structure. This work highlights a robust way to immobilize the highly mobile radionuclides with high waste loadings through encapsulation into durable matrices and a surface passivating mechanism that can greatly reduce the elemental transport from waste form materials and significantly enhance their chemical durability.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2020

Polymer–Ceramic Composite Electrolytes for Lithium Batteries: A Comparison between the Single-Ion-Conducting Polymer Matrix and Its Counterpart

Single-ion-conducting polymer electrolytes are attractive to use in lithium batteries as the transference number of the lithium cation approaches unity. This helps prevent concentration gradients across the electrolyte, which can result in dendrite formation. The addition of ceramic particles to polymer electrolytes at high loadings can increase the mechanical strength of the polymer, which can also help suppress dendrite formation. Here, a single-ion-conducting polymer electrolyte is blended with lithium-conducting oxide ceramic particles to make a composite electrolyte. This electrolyte is studied in comparison to a composite electrolyte containing freely dissolved lithium salt. It is found that the addition of ceramic particles to the single-ion-conducting polymer can result in increased cation dissociation and consequent increased ionic conductivity. The electrolytes are cycled in lithium symmetrical cells, and it is found that the ceramic-containing electrolytes show increased interfacial stability with the lithium metal compared to the pristine polymer electrolytes. Our findings shed light on how to optimize the polymer host chemistry to form composite electrolytes that can meet the challenging requirements to stabilize the lithium metal anode.

25 ENERGY STORAGE↗
DOE OSTI2026

Ceramic Composite Experimental Testing Status

Over recent years, ceramic matrix materials such as SiC–SiC and C–C have been gaining interest for use in fusion reactors, light water reactors (LWRs), and high-temperature reactors (HTRs). These materials are good candidates to operate in very high temperature and moderate to high radiation environments. The evaluation of composite materials, in general, is challenging because of variations in precursor materials, variations in the fabrication process across fabricators, and the wide range of potential fiber architectures, to name a few. However, the need to evaluate neutron-irradiated properties adds another layer of complexity, which includes cost, timeline, and specimen size limitations (often associated with irradiation testing). A qualification methodology for the use of ceramic composites is provided in the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Section III-5-HHB. The methodology is supported by ASTM International (ASTM) guides, which provide a pathway to accomplish this effort. Part of the qualification strategy is for the designer to collect material property data on environmental conditions representative of its design envelope. These data include irradiation effects. This report presents an experimental study and test campaign developed to partially address this gap by providing initial mechanical and physical property data required for design. A variety of different materials using different manufacturing techniques are considered as part of this campaign. The test plan suggests performing a screening or partial irradiation study to assist the designer during the material selection process. The designer can then perform a more comprehensive qualification study if the material performance is promising. This work focuses on the status of the specimen preparations (machining of samples), the current test methods and failure analysis as well as the preparation of irradiation vehicles for the irradiation campaign. The irradiation will be performed at Oak Ridge National Laboratory (ORNL) in the High Flux Isotope Reactor (HFIR) and at Idaho National Laboratory (INL) in the Advanced Test Reactor (ATR).

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Multi-stage load partitioning in additively manufactured Al6061+TiC nanocomposite characterized by in-situ neutron diffraction

Metal matrix composites (MMCs), combining metal matrix and ceramic particles, exhibit high mechanical strength and stiffness compared to conventional metallic materials. During fusion-based additive manufacturing (AM), MMCs undergo a rapid heating and cooling thermal history, which affects the interfacial bonding, thermal misfit stress and mechanical load transfer behavior between the metal matrix and particles, thus impacting the bulk mechanical performance. Here, we investigate the deformation dynamics and phase-specific load transfer in fusion-based AMed Al6061 +TiC MMCs through in-situ neutron diffraction. By calculating the phase-specific lattice strains, a multi-stage load transfer and deformation behavior during uniaxial compression is revealed: (1) elastic deformation in Stage I for both phases; (2) sudden stress rebalance between two phases in Stage II, evidenced by a sudden decrease in Al lattice strain and increase in TiC; (3) active load carrying by both phases in Stage III, with both phases experiencing an increase in lattice strains. Further, the deformation mechanism of each stage is deducted by correlating the evolutions of the interphase stresses and peaks’ broadening and intensity of the Al matrix. The local plastic deformation of the Al matrix near the phase interface is triggered and leads to stress rebalancing in Stage II. The global plastic deformation subsequently propagates throughout the Al matrix in Stage III, and the composite is further hardened through both dislocation multiplication and load transfer. The findings offer valuable insights into deformation and load-sharing behavior in AMed MMCs.

36 MATERIALS SCIENCE↗
DOE OSTI2025

A single-ion-conducting polymer and high-entropy Li-garnet composite electrolyte with simultaneous enhancement in ion transport and mechanical properties

Enabling the lithium metal anode has been the holy grail for improving the energy density for the next generation advanced batteries. Developing electrolytes that will suppress Li dendrite growth and provide sufficient ionic conductivity remains a major challenge in this field. In this study, we develop a polymer–ceramic composite electrolyte for lithium metal batteries. The polymer matrix is a vinyl ethylene carbonate (VEC) based single-ion-conducting polymer electrolyte. The ceramic filler is a Li 7 La 3 Zr 0.5 Nb 0.5 Ta 0.5 Hf 0.5 O 12 high-entropy Li-garnet (HE Li-garnet) ceramic, which is less prone to surface Li 2 CO 3 formation compared to Al-doped Li garnets. The addition of HE Li-garnet leads to a 7-fold increase in the ionic conductivity (8.6 × 10 −5 S cm −1 at 30 °C) compared to the pure polymer, while maintaining a high Li + transference of 0.73. Proton nuclear magnetic resonance and thermogravimetric analysis results suggest that the addition of HE Li-garnet results in a lower degree of polymerization of VEC, leaving more unpolymerized VEC monomers in the matrix, serving as the governing mechanism for conductivity enhancement. The favorable interactions between HE Li-garnet particles and the polymer matrix lead to a stable and well-mixed composite with 2-fold enhancement of storage modulus at 40 °C. The simultaneous ion transport and mechanical property enhancement significantly improves the composite electrolyte's dendrite resistance and cycle life in Li symmetric cells. This work highlights the positive role HE Li-garnet can play in improving polymer electrolytes to enable lithium metal anodes.

Ock, Ji-young [Oak Ridge National Laboratory (ORNL↗
DOE OSTI2023

Post-irradiation examination of the Sirius-1 nuclear thermal propulsion fuel test

Nuclear Thermal Propulsion (NTP) systems hold promise in reducing transit times for exploration of the solar system by both crewed and uncrewed missions. NTP systems currently under investigation include a once-through high-temperature gas-cooled fission reactor to provide thermal energy to heat the coolant, which also serves as the propellant. The fuel system consists of angular UN fuel particles dispersed in a matrix of W/Re, creating a ceramic and metallic composite or cermet, which has been irradiated in Idaho National Laboratory's (INL) Transient Reactor Test Facility (TREAT). This enabled evaluation of these materials under representative nuclear heating rates (~95 K/s) and peak surface temperatures (~2527 K). Further, these surface conditions were indicative of achievement of the target peak internal temperature of approximately 2850 K. These tests named Sirius-1 (UN–W/Re), have been irradiated and this paper will present post-irradiation examination results. The Sirius-1 test produced minor, stable cracks in the fuel specimen and spalling of surface material. Volatilized uranium ‘soot’ was found deposited on the inner wall of the irradiation capsule indicating loss of some fissile material from the fuel specimen. Spalling from the surfaces was also noted upon visual inspection. Uranium diffusion from the fuel particles resulted in the formation of U/Re phases and the production of a laminar microstructure on the edge of the fuel system. Overall, the specimen was stable and remained within a coolable geometry upon conclusion of multiple thermal cycles to prototypical operating temperatures.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗