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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 541 records · Page 30
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Damage progression and failure of SiC/SiC composite tubes under hard-contact radial expansion

The response of silicon carbide (SiC) fiber-reinforced SiC matrix (SiC/SiC) composite cladding to mechanical interaction with fissile fuel is a knowledge gap that must be overcome to design and assess SiC-based cladding systems for advanced nuclear applications. This study developed the relevant mechanical testing capability and identified the failure behavior and the critical microstructural features and processing defects. Sections of SiC composite tube were subjected to a modified expansion-due-to-compression (EDC) test in an X-ray computed tomography microscope: a polyurethane plug pressed surrogate Al 2 O 3 into the inner walls of the SiC/SiC composite tubes to achieve hard contact. A pure EDC test with just a polyurethane plug was also performed as a reference. Through the use of displacement fields, digital volume correlation revealed inhomogeneous deformation fields in the tubes, even for pure EDC, which was related to the inherent defects in the structure. Deep learning–aided segmentation and systematic data analysis revealed that the presence of inhomogeneous deformation applied by the hard contact was exaggerated by the presence of inner surface imperfections left behind from the matrix densification process. In conclusion, the findings provide insights into the applications, highlighting the necessity for improvements in inner surface roughness and the incorporation of localized contacts in pellet–cladding mechanical interaction computational models.

Composites↗
DOE OSTI2025

Genesis of a novel high-rate composite manufacturing process using large-scale additive manufacturing – compression molding (AM-CM) system: Possibilities and limitations

Oak Ridge National Laboratory (ORNL) has developed a highly automated manufacturing process for thermoplastic composites that combines the benefits of Additive Manufacturing and Compression Molding (AM-CM) to produce high-performance functional composite structures at automotive production rates. Here, the AM-CM process creates highly precise preforms by additively placing extruded fiber-filled polymers (with controlled fiber orientations and multi-material configurations) in the desired mold location before undergoing a secondary compression molding process immediately before the preform cools down. Preforms can be in the form of short, long-chopped, or continuous fiber-filled thermoplastic polymers (e.g., CF/GF-filled ABS, PC, LM-PAEK, etc.). The AM-CM process combines the benefits of controlled fiber alignment, that is only achievable in AM-printed parts with the classical CM process, which eliminates porosity and good surface finish. A preform created using AM-CM can integrate various materials to enable additional architectural functionalities, including over-molding, selective stiffening, and the incorporation of electrically or thermally conductive channels. All these advantages come with a fast part production cycle time. The AM-CM process can manufacture multi-material, multi-functional parts in under 3 min, starting from raw material (pellets) to the final product. The novel AM-CM process offers superior microstructural control and enhanced multi-functionality previously unattainable with any other traditional high-rate thermoplastic composite manufacturing method. This work covers the manufacturing concept, system development, materials and applications of AM-CM process in detail.

Kumar, Vipin [Oak Ridge National Laboratory (ORNL)↗
DOE OSTI2025

TRANSP integrated modeling code for interpretive and predictive analysis of tokamak plasmas

This paper provides a comprehensive review of the TRANSP code, a sophisticated tool for interpretive and predictive analysis of tokamak plasmas, detailing its major capabilities and features. It describes the equations for particle, power, and momentum balance analysis, as well as the poloidal field diffusion equations. The paper outlines the spatial and time grids used in TRANSP and details the equilibrium assumptions and solvers. Various models for heating and current drive and radiation, including updates to the NUBEAM model, are discussed. The handling of large-scale events such as sawtooth crashes and pellet injections is examined, along with the predictive capabilities for advancing plasma profiles. The integration of TRANSP with the ITER Integrated Modeling and Analysis Suite (IMAS) is highlighted, demonstrating enhanced data access and analysis capabilities. Additionally, the paper discusses best practices and continuous integration techniques to enhance TRANSP's robustness. The suite of TRANSP tools, designed for efficient data analysis and simulation, further supports the optimization of tokamak operations and coupling with other tokamak codes. Continuous development and support ensure that TRANSP remains a major code for the analysis of experimental data for controlled thermonuclear fusion, maintaining its critical role in supporting the optimization of tokamak operations and advancing fusion research.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2026

A multiscale packed-bed reactor model for sustainable ethylene production via chemical looping oxidative coupling of methane

The rising global warming concerns and shale gas discovery have prompted research in the direction of greenhouse gas (GHG), such as methane, reduction and conversion. Oxidative coupling of methane (OCM) offers a pathway to low carbon-intense valorization of methane while producing ethylene, a chemical regarded as central to the petrochemical industry. Even after decades of OCM discovery, researchers keep understanding the process and underlying chemical reactions in a pursuit to achieve industrial viability for OCM. Here, in general, OCM suffers from low C 2 selectivity, yield and reactor temperature runaways due to highly exothermic nature of its reactions. Computational Fluid Dynamics (CFD) tools help analyze spatial gradients within the reactor to deeply understand the diffusion of species, mass and heat transfer phenomena. Furthermore, challenges associated with scaling up such as hot spot formation and parametric sensitivity can be addressed without having to expend on costly experiments. The current paper presents a multiscale packed-bed reactor CFD model coupled with a chemical kinetic model for the chemical looping OCM. The CFD model includes two scales i.e., macroscale for catalyst bed and microscale for individual pellets. Moreover, a chemical kinetic model based on 10 gas-phase reactions is integrated with the CFD model. An additional surface reaction for the formation of gas-phase oxygen from catalyst surface is added to account for the absence of feed oxygen. The model is calibrated against experimental results. The calibrated model captures trends in CH 4 conversion, C 2 selectivity and C 2 yield within a ± 4.35 % range across a temperature range of 700-900 °C. Moreover, model fidelity is evaluated by varying key computational parameters such as mesh resolution and time step size. The model is also verified by varying the inlet methane concentration and the gas hourly space velocity (GHSV) and comparing the results with literature. A sensitivity analysis and scale-up of the current model is undergoing.

Chemical looping↗
DOE OSTI2026

Technoeconomic Analysis of Steel Production with Electric Thermal Energy Storage

The iron and steel industry is an important manufacturing sector and one of the largest energy consumers in the United States and globally. Hydrogen direct reduction of iron ore (H2DRI) is considered a promising process that could enhance domestic steel production. This process requires hydrogen inlet temperatures up to 950 degrees C to drive the endothermic reduction of the iron ore pellets. In this work we investigate the technoeconomic performance of an H2DRI plant using electric thermal energy storage (ETES) technologies for the hydrogen heating, compared to conventional natural gas fired heaters, hydrogen fired heaters, and electric hydrogen heaters. A technoeconomic analysis framework for the plant is developed and used in multiple case studies, covering different hydrogen prices, grid electricity profiles, and financing scenarios. The levelized cost of steel production is found out to be in the range of $775-950/mt, which is mostly inside the benchmarked steel price of $941/mt. ETES-based hydrogen heating is found to be in par with conventional natural gas fired heaters, and cheaper than hydrogen fired heaters and electric hydrogen heaters. The major cost drivers are the iron ore and hydrogen feedstock, followed by the hydrogen compression and heating capital. Several insights and suggested future directions are identified.

08 HYDROGEN↗
DOE OSTI2021

Assessment of the burning-plasma operational space in ITER by using a control-oriented core-SOL-divertor model

In future tokamaks, the control of burning plasmas will require careful regulation of the plasma density and temperature. Along with the design of effective burn-control systems, understanding how the fusion power varies in the density-temperature space is vital for the operation of fusion power plants. Here in this work, the steady-state operational space of ITER is studied using a control-oriented core-plasma model coupled to a two-point model of the scrape-off-layer (SOL) and divertor regions. The two models are coupled through the exchange of input-output parameters. The deuterium and tritium recycling from the wall are output parameters of the SOL-divertor model that are used as input parameters in the core-plasma density balance. Furthermore, the separatrix temperature, which is an output parameter of the SOL-divertor model, is incorporated into the radial core-plasma temperature profiles. Therefore, the temperature-dependent power balance of the plasma core is intimately linked to the SOL-divertor model. Both the power entering the SOL from the core, as determined by the core-plasma power balance, and the separatrix density, as dictated by the core-plasma density balance, are input parameters to the SOL-divertor model. They are control knobs in the SOL-divertor model that can be regulated using the core-plasma actuators: auxiliary power and pellet injection. There are various operational limitations, such as the saturation of the aforementioned actuators, that will prevent ITER from accessing certain high-fusion plasma regimes. The achievable tritium concentration in the fueling lines and the maximum sustainable heat load on the divertor will impose further restrictions. By accounting for these limitations, the ITER operational space is computed based on the coupled core-SOL-divertor model and visualized using Plasma Operation Contour (POPCON) plots that map performance metrics, such as the fusion to auxiliary power ratio, over the density-temperature space. Comparisons are drawn between plasmas with different recycling, confinement, and SOL-divertor conditions.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

Control-oriented core-SOL-divertor model to address integrated burn and divertor control challenges in ITER

The real-time regulation of a burning plasma’s temperature and density, or burn control, will be necessary to produce high fusion power in future tokamaks like ITER. This is made more challenging due to the plasma’s nonlinear characteristics and the interdependence between the core-plasma and edge-plasma regions. For example, a raising plasma temperature leads to increasing reactivity and therefore to more alpha-particle heating, which further increases temperature. Furthermore, a raise of the fusion power increases the heat flow through the scrape-off-layer (SOL), which can compromise the integrity of the divertor without proper safeguards. For control design, a model-based approach is attractive because it can directly incorporate the nonlinear, coupled, burning-plasma dynamics into the design. To facilitate this design approach, a control-oriented core-SOL-divertor (CSD) model is presented in this work. In this CSD model, a core-plasma model captures the nonlinear dynamics of the core’s density and temperature, and a SOL-divertor model defines the plasma conditions at the separatrix and divertor including the heat load on the target plates. The core-plasma and SOL-divertor models are coupled through the exchange of various variables. In particular, the SOL-divertor model yields the separatrix temperature and the influx of recycled particles into the core-plasma. Further, these variables influence the power and particle balances captured by the core-plasma model. In return, the core-plasma model determines the intensity of the heat and particles fluxes across the separatrix, and this outflow strongly impacts the SOL-divertor model. Therefore, the power and density of the core-plasma, which can be readily modulated through external heating systems and pellet injection, can be viewed as control knobs for the SOL-divertor region in addition to the gas puffing. In simulations of the CSD model, it is demonstrated how external actuation can be utilized to meet burn control and divertor control objectives simultaneously.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2020

Aluminum-doped U 3 Si 2 composite fuels with enhanced oxidation resistance

Al-doped U 3 Si 2 composite fuels with controlled microstructure were fabricated by spark plasma sintering that display greatly-improved oxidation resistance as compared with monolithic and Al-doped silicides prepared by standard powder metallurgy or arc melting. The effects of Al additives on the thermal-mechanical properties and oxidation resisance of the micron- and nano-sized U 3 Si 2 composites were investigated. Additionally, a minimal addition of 1.8 at% Al is effective to increase the onset oxidation temperature of as-fabricated U 3 Si 2 pellets to 580 °C, which can be further increased to 610 °C by thermal annealing. The Al-doped U 3 Si 2 composite fuels also display simultaneously higher hardness and fracture toughness than un-doped U3Si2. These results highlight an effective strategy by integrating minimal Al additives, microstructure control and post-thermal annealing to design advanced silicide fuels with excellent oxidation resistance, desired thermal-mechanical properties and maintained high fissile element density.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Compositional screening of Ce-doped (Gd,Lu,Y) 3 (Al,Ga) 5 O 12 ceramics prepared by quenching from melt and their luminescence properties

Quenching from the melt using an Optical Floating Zone furnace was investigated as a possible high-throughput preparation method in order to screen novel scintillating materials. To validate this method, polycrystalline rare-earth aluminum garnets and yttrium gallium aluminum garnets were synthesized and characterized by X-ray diffraction, photoluminescence and radioluminescence emission spectra as well as radioluminescence imaging and compared to the previously reported properties of those materials. In order to achieve rapid, but quantitative comparison of the materials, pellets fabricated from the synthesized powders were sintered; two different sintering conditions were investigated and compared. Therefore, a simplistic energy deposition and light absorption model for brightness of opaque samples under X-ray irradiation was developed to estimate the relative scintillation efficiency of the materials. Based on the results, Y 3 Al 5 O 12 , LuY 2 Al 5 O 12 and GdY 2 Al 5 O 12 seem to have the highest scintillation efficiency among the prepared samples.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Densification of non-radioactive porous siliceous particles loaded with cesium potassium copper hexacyanoferrate by spark plasma sintering

In the development of a conditioning process for cesium-bearing sorbents from the decontamination of aqueous waste streams, the treatment of a granular material composed of (Cs,K) 2 CuFe(CN) 6 loaded onto mesoporous silica particles has been investigated in an additive-free two-step conditioning process involving (i) calcination at 1000 °C to remove volatile species and (ii) densification by spark plasma sintering. A 92.0 % dense pellet was obtained after 10 min sintering at 850 °C under 100 MPa. Density and microstructural measurements indicate that densification is first driven by the viscous flow of a vitreous phase containing Cs, formed during calcination. Additionally, the density of the sample then increases further during sintering by grain boundary diffusion. However, a limited but non negligible part of cesium is lost during densification.

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 OSTI2020

In-pile thermal conductivity of uranium dioxide at low burnup

In-pile thermal conductivity of uranium dioxide (UO 2 ) was investigated at low burnup levels (<0.2 kWd/kg-UO 2 ) to elucidate the concurrent effects of the fission damage and the thermal recovery. In-pile experiments were performed in the Halden reactor. The fuel centerline and the cladding temperatures were measured during the experiments, and the power level of specially designed test rodlets were also monitored. The uncertainty of the pellet-cladding gap's thermal resistivity was minimized with a liquid-metal bond, while comparable measurements were made in reference test rodlets without liquid metal bonding. The experimental data were analyzed using the inverse heat transfer approach. The heat conduction equation was solved by applying the measured temperature of the cladding as a boundary condition, and the calculated fuel centerline temperature was compared with the measured fuel centerline temperature to determine the relative matrix resistivity. Results showed that the in-pile matrix resistivity, the phonon-lattice interaction terms of the thermal conductivity, was ~1.5–2.5 times higher than its unirradiated value due to fission-induced damage. Finally, the in-pile annealing caused a significant recovery of the matrix resistivity, and the amount of recovery increased from ~20 to ~33%, while the annealing temperature increased from 700 to 1,000 °C.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Phase-field simulations of intergranular fission gas bubble behavior in U 3 Si 2 nuclear fuel

U 3 Si 2 is a potential accident-tolerant fuel that shows promise due to its high thermal conductivity and higher uranium density relative to UO 2 . However, its swelling and fission gas release behavior in light water reactor (LWR) conditions is relatively unknown. To provide mechanistic insight and determine parameters for engineering-scale fuel performance modeling of pellet-form U 3 Si 2 , phase-field simulations of the growth, interconnection, and venting of intergranular fission gas bubbles were performed. The fractional coverage of the grain boundary and the fraction of bubble area that is vented were calculated as a function of time. From the simulation data, the fractional grain boundary coverage at saturation, an important parameter needed in engineering-scale modeling of swelling and fission gas release, was determined. Multiple simulations were run to determine the uncertainty in the calculated value. Finally, the effect of model assumptions and input parameters that are not well known was evaluated. Simulation results are compared to related theoretical and computational work. Based on the simulation results, a value of 0.60 for the fractional grain boundary coverage at saturation is recommended for U 3 Si 2 fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

3Y-TZP Toughened and Oxidation-resistant U 3 Si 2 Composites for Accident Tolerant Fuels

In this work, we report an innovative approach synergizing multiple effects of microstructure control, oxide protection and phase transformation-induced mechanically toughening to develop advanced U 3 Si 2 fuels with state-of-the-art materials properties. 3Y-TZP (3 mol% yttria doped stabilized tetragonal zirconia) additives are incorporated and uniformly distributed into the SPS-densified U 3 Si 2 fuel matrix, enhancing materials fracture toughness up to 4.04 MPa m 1/2 . The 3Y-TZP incorporated U 3 Si 2 composite pellets show significantly improved onset temperature of oxidation above 560 °C, which can be further improved to 617 °C with 3 vol% addition by post-sintering thermal annealing. We report the development of mechanically tough and oxidation-resistant U 3 Si 2 with minimal additive represents a major step forward towards realizing the potential of high density U 3 Si 2 as the leading fuel concept to increase accidence tolerance of nuclear energy systems.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Small uranium and oxygen interstitial clusters in UO 2 : An empirical potential study

The aggregation of defect clusters in UO 2 under irradiations creates dislocation loops and cavities, which in turn contribute to swelling of the fuels, leading to detrimental pellet-clad mechanical interactions. The kinetics of the defect clusters, however, have received less attention and are not well understood. As such, using two interatomic potentials, the Morelon potential and a many-body model by Cooper, Rushton, and Grimes (CRG), the energetics and kinetics of small uranium interstitial clusters, with and without bound oxygen interstitials, are investigated with molecular dynamics and accelerated molecular dynamics simulations. For the single bound anti-Schottky interstitial cluster (U i 2O i ), both potentials find the interstitialcy migration mechanism to be a low barrier migration mechanism. However, for the CRG potential, an oxygen Frenkel pair assisted interstitial cluster migration mechanism with a lower migration barrier is identified during accelerated molecular dynamics simulations. For the Morelon potential, the interstitial clusters have geometrically compact structures, and small interstitial clusters (two and three anti-Schottky clusters) are more mobile than the single bound anti-Schottky interstitial cluster; for the CRG potential, the interstitial clusters have geometrically non-compact structures, yielding a more complex energy landscape. The rapid migration of interstitials and interstitial clusters are important for the irradiation response of UO 2 and correlates well with the very low barriers reported in a few historic experimental studies.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2021

Electron microscopy characterization of the fuel-cladding interaction in medium burnup annular fast reactor MOX

In this work, we present electron microscopy data focused on the fuel-cladding interaction layer in annular fast reactor MOX with HT-9 cladding at medium burnup. In agreement with previous literature data, the volatile fission products Cs, Te and I have migrated radially into the extreme fuel periphery and partially interacted with the cladding. The accumulation of Cs has occurred in the outermost rim of the fuel pellet, where grain recrystallization has also been observed. Significant amounts of Pd have been found in the interaction zone, particularly in the sample taken from the upper half of the fissile column where the cladding temperatures are higher. At this axial location, Cr has been enriched at the cladding inner surface and diffused into the fuel. Furthermore, chromium remains mainly in metallic form, but locally formed oxides. The fission products Cs, Te and I are found, with variable composition, in form of nanocrystalline regions dispersed in the metallic Cr-rich layer. The morphology and chemical characteristics of the layer suggest a non-oxidative corrosion mechanism as principal cladding degradation phenomenon occurring in this sample, with local onset of Cr oxidation within the nanocrystalline precipitates.

36 MATERIALS SCIENCE↗