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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 37 records · Page 2
All sourcesNASA NTRSDOE OSTI
NASA NTRS1984

Chemical feasibility of lithium as a matrix for structural composites

The chemical compatibility of lithium with tows of carbon and aramid fibers and silicon carbide and boron monofilaments was investigated by encapsulating the fibers in liquid lithium and also by sintering. The lithium did not readily wet the various fibers. In particular, very little lithium infiltration into the carbon and aramid tows was achieved and the strength of the tows was seriously degraded. The strength of the boron and silicon carbide monofilaments, however, was not affected by the liquid lithium. Therefore lithium is not feasible as a matrix for carbon and aramid fibers, but a composite containing boron or silicon carbide fibers in a lithium matrix may be feasible for specialized applications.

Swann, R. T.↗
NASA NTRSDate not supplied

Additive Manufacture of Novel Tungsten Wicking and Heat Transfer Structures

Additive manufacturing (AM) is utilized for fabricating complex geometries from refractory metals, such as tungsten and molybdenum. In this study, plates with ultra-fine and fine lattice structures were manufactured from tungsten to demonstrate printability and the component fabrication advantages of AM. Previous work was completed on individual tungsten lattice structures; this work demonstrates the advantages of printing these structures on tubing and first-wall materials. Initial work using stainless steel demonstrated increased heat transfer properties promoted by these lattice structures. Applications for these structures include wicking structures, heat pipes, and first-wall materials in fusion devices. A major advantage of these designs is the capability to produce a monolithic component with tailored heat transfer and wicking properties. This research helps promote the ability to use liquid lithium in fusion devices since these lattice structures manufactured from tungsten could act as a high temperature first-wall or divertor material and as a wicking structure for liquid lithium. Different lattice structures with varying unit cell sizes and strut thicknesses were printed and the effects of these structures on heat transfer properties were measured. Abrasion resistance of these structures was also investigated. This presentation will discuss the results of these experiments and their implications related to heat transfer applications. This work will help inform the development and implementation of complex geometries made from tungsten for future heat pipe and fusion first-wall or divertor component designs.

Carly J. Romnes↗
DOE OSTI2024

NSTX-U National Research Program: White Paper in Response to Call from FESAC Sub-Committee

Both scientific and technical innovation is needed for the realization of an attractive engineering solution for a timely and cost-effective Pilot Plant, the design and construction of which is the overarching recommendation of the FESAC Long Range Plan, and the 2021 NASEM Pilot Plant reports, which underpin the Bold Decadal Vision. The two most significant plasma physics gaps to close for a Compact Pilot Plant (CPP) are core confinement improvement and heat flux mitigation, neither of which have been closed in an integrated fashion for any planned fusion power production device. High core confinement and stability are essential for producing majority self-driven plasmas in CPPs with reduced size and auxiliary heating power requirements, with an improvement in confinement being the major driver for cost reduction of a CPP. The National Spherical Tokamak Experiment - Upgrade (NSTX-U) is a unique low aspect ratio research facility that will address the fundamental challenge of developing the science and technology basis for a CPP design that integrates high core and edge confinement with the ability to mitigate very high incident heat fluxes. NSTX-U capabilities will enable the high performance, already achieved on NSTX, to extend into physics regimes much closer to those anticipated in Spherical Tokamak (ST)-based CPPs. These confinement and stability properties will be assessed by a full complement of diagnostics and analysis tools, which will also aid in the development of the underlying theory and predictive models needed for further optimization. Both conventional and transformative heat flux mitigation methods, such as liquid lithium plasma-facing components, will be developed and tested in-situ in NSTX-U at incident heat fluxes of ~100 MW/m 2 , and will inform plans and reduce risk for a subsequent major upgrade to the device to fully heated, high-Z wall and full liquid lithium divertor capability, a technology that potentially could then be implemented on any magnetic confinement device at any aspect ratio. NSTX-U research is fully complementary to programs performed on other STs, nationally and internationally. Furthermore, NSTX-U research has a direct connection to the private sector by informing design choices for future power production facilities being developed by these companies. The NSTX-U program will operate as a national User Facility, with collaborating researchers, engineers, and graduate students from 19 outside institutions, and open to participation and experiments led by researchers from both public and private entities. The research program will advance workforce development through training of young scientists, engineers, and technicians, and it will also serve for further diagnostic innovation, especially for high heat flux and high-Z wall environments, and implementation of advanced artificial intelligence (AI) for plasma and heat flux control.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
NASA NTRS1971

Generation of long time creep data on refractory alloys at elevated temperatures

Ultrahigh vacuum creep tests were conducted on tantalum, tungsten, and molybdenum alloys to develop creep data and to evaluate the influence of liquid lithium exposure on the creep behavior of the tantalum alloy T-111. Test conditions were generally selected to provide 1% creep in 1000 to 10,000 hours, with test temperatures ranging from 1600 to 2912 F (1144 to 1873 K). Tests on T-111 specimens exposed to vacuum and to liquid lithium for various times and temperatures showed large reductions in creep strength for exposures in the 1800 to 1900 F (1255 to 1310 K) range. Tests on the ASTAR 811c showed a significant influence of grain size on the creep strength of this material, with larger grain size specimens being significantly stronger above the equicohesive temperature of 2000 F (1366 K).

Sheffler, K. D.↗
NASA NTRS1951

Theoretical performance of lithium and fluorine as a rocket propellant

Theoretical performance for liquid lithium and liquid fluorine as a rocket propellant was calculated with assumptions both of equilibrium and frozen composition during expansion. Parameters included were specific impulse, combustion-chamber temperature, nozzle-exit temperature, composition, mean molecular weight, characteristic velocity, coefficient of thrust, and ratio of nozzle-exit area to throat area. For chamber pressure of 300 pounds per square inch absolute and expansion to 1 atmosphere, the maximum equilibrium specific impulse calculated was 335.5 pound-seconds per pound. The effect of ionization on calculated performance was shown to be negligible by comparison of values of various parameters calculated both with and without ionized products of combustion.

FUELS -ROCKETS (INCLUDES FUELS AND OXIDANT)↗
DOE OSTI2024

NSTX-U liquid metal core-edge facility (LMCE)

NSTX-U/LMCE will provide a unique and world-leading research facility to address the primary challenge to delivering economic and timely magnetic fusion energy, namely the need to develop a power and particle exhaust and first-wall system that can withstand very high edge heat fluxes, maximize energy confinement, and avoid the production of large masses of solid eroded first-wall material. The NSTX-U/LMCE facility will assess the ability of liquid metals (LMs) – especially liquid lithium – to provide a new boundary condition for magnetic fusion systems, to extend the lifetime of the plasma facing components (PFCs) and improve core plasma confinement. Such capability is needed to establish the basis for next-step fusion facilities including fusion pilot plants, and to maintain U.S. world leadership in core-edge integration research. NSTX-U/LMCE will leverage the ability to generate very high divertor perpendicular heat flux q⊥ ~ 100MW/m 2 , extensive diagnostics, and liquid-metal-applicable infrastructure of NSTX-U. NSTX-U/LMCE will provide access to a high-confinement plasma core with majority self-driven plasma current, the flexibility to test a range of liquid metal divertor concepts, access to a range of separatrix collisionalities (from high to very low), and the ability to controllably vary the first-wall temperature to vary the plasma- wall interaction physics on liquid lithium components. Further, NSTX-U/LMCE will utilize more reactor-relevant high-Z refractory-metal PFC substrates. With these capabilities the NSTX-U/LMCE facility will explore the full continuum of core-edge solutions ranging from high core radiated power, to conditions with radiative losses concentrated in the scrape-off layer (SOL), and ultimately low recycling conditions. The low collisionality SOL that may be accessible in the low recycling regime is relatively unexplored and will require a kinetic treatment of the edge, which can be addressed theoretically, and with experiments in LTX-β. Additional smaller-scale preparatory R&D facilities will be required to reduce the risk of premature technical/engineering failure of liquid metal systems implemented in NSTX-U. The NSTX-U/LMCE facility aligns very well with recommendations in the FESAC Long-Range Plan and NASEM Pilot Plant reports and the Bold Decadal Vision, will be unique in the world program throughout the next decade, and is garnering private company interest in utilizing NSTX-U/LMCE for development of LM PFCs.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
NASA NTRS1976

Corrosion of oxygen-doped tantalum by lithium

The increase in the penetration of tantalum by liquid lithium with increasing oxygen content of tantalum results from the ability of lithium to extract oxygen from tantalum oxides and/or from regions of high oxygen concentrations, accompanied by the creation of minute channels. At low oxygen concentration, oxides are present primarily at the grain boundaries and the lithium penetration is primarily intergranular. At high oxygen concentration, oxides and/or regions of high oxygen concentration are present both at the grain boundaries and within the grains; lithium penetration is both intergranular and transgranular. For a specific oxygen concentration, the severity of corrosion decreases with increasing temperature because (1) the oxygen solubility increases with temperature, decreasing the number of oxide particles and/or regions of high oxygen concentration, and (2) the diffusivity of oxygen in tantalum increases, allowing more rapid oxygen removal at the tantalum lithium interface.

Stecura, S.↗
DOE OSTI2026

INFUSE:FLARED (Final Technical Report)

This report summarizes the final outcomes of the INFUSE:FLARED collaboration between Tokamak Energy (TE), the University of Illinois Urbana-Champaign (UIUC), and the Department of Energy (DOE). The project aimed to experimentally determine and model the pumping, absorption, and separation behavior of deuterium in flowing liquid lithium under gas and plasma exposure conditions. The work supports the development of lithium-based plasma-facing components (PFCs) and tritium management strategies for next-generation fusion plants.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
NASA NTRS1972

Thermochemistry of tantalum-wall cooling system with lithium and sodium working fluids

Plots are presented which show the distribution of oxygen between liquid lithium and tantalum or niobium, and between liquid sodium and tantalum at elevated temperatures. Additional plots showing the composition of the gas phase above the solutions of oxygen and alkali metal are presented. The use of the plots is illustrated by an example tantalum heat pipe filled with lithium.

Tower, L. K.↗
NASA NTRS1973

Preparation of refractory cermet structures for lithium compatibility testing

High-purity nitride and carbide cermets were synthesized for compatability testing in liquid lithium. A process was developed for the preparation of high-purity hafnium nitride powder, which was subsequently blended with tungsten powder or tantalum nitride and tungsten powders and fabricated into 3 in diameter billets by uniaxial hot pressing. Specimens were then cut from the billets for compatability testing. Similar processing techniques were applied to produce hafnium carbide and zirconium carbide cermets for use in the testing program. All billets produced were characterized with respect to chemistry, structure, density, and strength properties.

Heestand, R. L.↗
DOE OSTI2024

Optimization of lithium vapor box divertor evaporator location on NSTX-U using SOLPS-ITER

Commercial fusion reactors will be faced with extremely high divertor target heat fluxes that will require mitigation. Simulations of detachment in an NSTX-U scenario projected to have 92 MW/m 2 unmitigated peak target heat flux are presented, which reaches sub-10 MW/m 2 target heat flux using a highly dissipating lithium vapor box divertor design. The lithium vapor box is a detached divertor design which employs lithium vapor evaporation and condensation to contain lithium below the X-point. Previous SOLPS modeling has indicated a lithium vapor box can reduce the heat flux down to 10 MW/m 2 via simultaneous evaporation from the Private Flux Region (PFR) and the Common Flux Region (CFR) sides of the vapor box. It is found here that PFR evaporation has improved access to the separatrix leading to significantly more efficient power dissipation than CFR evaporation. Simulations of target evaporation with an evaporation distribution that is self-consistent with the temperature of a Capillary Porous System with Fast flowing liquid lithium (CPSF) could reach n Li / n e ~ 0.025-0.030 at the Last Closed Flux Surface (LCFS) depending on the liquid metal flow speeds and lithium sputtering yield, while PFR-side evaporation can reach acceptable heat fluxes with n Li / n e ~ 0.038 at the LCFS. However, PFR evaporator performance can be improved if the target is allowed to be hot enough such that it reflects lithium, reaching n Li / n e ~ 0.028 and reducing required lithium evaporation. Ultimately PFR evaporation and target evaporation are found to have similar ability to produce acceptable heat flux solutions with minimal upstream concentration.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2026

Effects of negative triangularity on microinstabilities in a low-recycling lithium-wall spherical tokamak

In this work, we present a linear gyrokinetic study of the impact of negative triangularity (NT) on microinstabilities in the Lithium Tokamak eXperiment-β (LTX-β), a low-recycling spherical tokamak with liquid lithium plasma-facing components that produce flat electron-temperature profiles [Elliott et al., IEEE Trans. Plasma Sci. 48, 1382 (2020)]. While NT is widely recognized as a stabilizing mechanism and often associated with improved confinement in conventional tokamaks, this study reveals that its effect is not universally stabilizing in the parameter regime of LTX-β and is shown to be highly sensitive to local equilibria. Using local linear simulations with the GS2 code [Kotschenreuther et al., Comput. Phys. Commun. 88, 128 (1995)] at ρ=0.3, 0.5, and 0.8 for two representative discharges (#103955 and #109355), and employing the Miller equilibrium model to isolate shaping effects, we find that NT can transition from stabilizing to destabilizing depending on magnetic shear, safety factor, and electron-temperature gradient. In shot #103955, NT reduces growth rates across radii, with strongest stabilization at the edge, whereas in shot #109355, it is stabilizing in the core but destabilizing at mid-radius and edge under experimental conditions. Parametric scans show that flattening the electron-temperature profile, increasing magnetic shear, and reducing the safety factor recover NT stabilization. These results demonstrate that NT stabilization is tunable rather than intrinsic and requires coordinated control of magnetic geometry and gradient drive.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
NASA NTRS1996

Aerospace Gas/Liquid Separator for Terrestrial Applications

The space gas/liquid separator, a key component in the heat transport subsystem of a space reactor power system, was developed to remove helium gas from liquid lithium in zero gravity. This application is opposite from the space application in that it is removing a liquid from a gas rather than a gas from a liquid.

space↗
DOE OSTI2025

Self-consistent modeling of tokamak edge plasma transport with lithium sources

Magnetic confinement fusion devices require effective heat and particle exhaust solutions on the divertor plates to operate sustainably, especially under reactor-relevant conditions. Liquid lithium divertors have been proposed to address two major challenges: control of excessive heat flux to plasma-facing components through vapor shielding and minimization of core plasma contamination from impurities. The National Spherical Torus Experiment-Upgrade (NSTX-U) will explore lithium as a divertor material due to its potential to meet both objectives. We present a self-consistent coupling framework between the plasma boundary transport code UEDGE and the lithium wall transport code Wall–Li to evaluate the feasibility and operational limits of lithium-based divertors. The model aims to optimize lithium sourcing levels to prevent core plasma contamination via fuel dilution while ensuring divertor protection through vapor shielding. This integrated framework, applicable to any tokamak with lithium sources, dynamically adjusts lithium sourcing based on local plasma conditions and surface temperature. The coupled model is tested using NSTX-like geometry and plasma conditions to assess its performance and reliability. Wall–Li calculates lithium fluxes from plasma-facing components, incorporating physical sputtering, thermally enhanced sputtering, and evaporation driven by surface temperature and ion flux. These fluxes are reintroduced into UEDGE as neutral lithium atoms, enabling simulation of their transport and distribution within the plasma. UEDGE computes plasma and neutral transport, surface heat flux, and iteratively feeds this information back to Wall–Li. A small time step is employed to ensure numerical stability and convergence, enabling accurate simulations over typical tokamak discharge durations. This integrated modeling approach provides a robust tool for identifying operational regimes that balance effective lithium sourcing with minimal core plasma contamination, offering critical insights for optimizing lithium-based divertor systems in current and future fusion devices.

Magnetic confinement fusion↗