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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 OSTI2024

Net lithium deposition and dominant self-sputtering in lithium tokamak experiment-β with a liquid lithium wall

We observed enhanced net lithium deposition and lithium erosion, possibly dominated by physical sputtering of lithium by lithium-ion bombardment, on the outer plasma-facing surface in the Lithium Tokamak eXperiment-β (LTX-β) during liquid lithium wall operations. Silicon crystal samples with micro-trenches (30 μm × 30 μm × 2–7 μm deep) were exposed to hydrogen plasmas in LTX-β for solid and liquid lithium wall operations. Post-mortem analysis using X-ray photoelectron spectroscopy combined with argon ion sputtering measured net lithium deposition of 8.2 or 21 nm on the silicon crystal surface exposed for ~ 50 repeated shots of ~ 50-ms hydrogen plasma discharges during the liquid lithium wall operations at a vessel temperature of 475 K. Energy dispersive X-ray spectroscopy measured oxygen concentration patterns on the micro-trench floors, which were due to oxidized lithium deposition. Using the inhomogeneous oxygen concentration pattern caused by an ion-shadowing effect associated with the micro-trench’s geometric structure, we determined a polar incident ion direction of 68.4 ± 1.6° referenced to the surface normal direction. This observation was well-explained by the hypothesis that self-sputtering of Li was a dominant lithium erosion source in addition to lithium sputtering by hydrogen bombardment.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

Improved liquid lithium surfaces in the Lithium Tokamak Experiment-β

Advances in vacuum, surface, and lithium conditioning techniques throughout five years of continuous operations in LTX-β have produced mirror-like liquid lithium surfaces and demonstrated the feasibility of high-performance tokamak discharges fully surrounded by liquid metal without significant operational problems. Improvements in conditioning techniques and procedures, including many weeks of baking and accumulation of 70 g of Li, led to reduced residual gasses and clean Li surfaces - all while still maintaining enough operational flexibility for multiple in-vacuum diagnostic upgrades and calibrations. Coatings had a visibly clean appearance, with reflective liquid metal demonstrating good wetting and surface adhesion with films that were now macroscopically thick. Solidified Li showed large crystal grains, while surface science measurements observed reduced impurities in the lithium. Steadily improved plasma performance was achieved with liquid lithium, with discharges able to match solid Li in terms of evolution of I p and n e , including rapid density pumping indicating low recycling. There were indications of moderately increased Li impurity influx, though few significant disturbances by the large liquid surfaces on tokamak operations over hundreds of discharges. Liquid metal plasma facing components are a potential solution to the extreme heat and particle fluxes that could cause unacceptable damage to solid materials, while liquid lithium also has the potential for greatly increased confinement in the low-recycling regime. While many liquid metal approaches are possible, and numerous experiments have been conducted in test stands and small modules in fusion devices, LTX-β is the only tokamak operated while fully surrounded by liquid metal.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Liquid lithium divertor analysis using coupled plasma material interaction model

A liquid lithium divertor can improve performance of future fusion devices by creating efficient power exhaust and improving the energy confinement via pumping of the hydrogen isotopes. In addition, significantly higher heat fluxes can be handled if controlled vapor shielding is used to redistribute the divertor heat flux over a wider area. Design and optimization of such a system calls for an analysis model which includes a strong two-way coupling between the plasma and divertor material. The incoming plasma heat and particle flux will affect the divertor surface temperature, which is a defining factor of the lithium evaporative and sputtered flux going into the plasma. Results of the coupled model based on the plasma edge code SOLPS-ITER and the computational fluid dynamics (CFD) code ANSYS-CFX will be presented for different configurations. An analytical slab flow model is used as a heat transfer boundary condition for SOLPS, defining particle flux from the wall via calculation of the surface temperature. At the final step, results of the SOLPS analysis are verified using a 3D CFD magnetohydrodynamics (MHD) analysis which uses heat and particle flux from SOLPS as a boundary condition. In addition to plasma heat flux, both analytical and CFD temperature models include several plasma material interaction effects, such as lithium evaporation, condensation and sputtering based on deuterium target flux. New adatom sputtering model based on the available experimental data is presented. Analytical model is expanded to include free surface axisymmetric configurations. Results of parametric studies of the divertor configurations with different lithium inlet temperature and velocity will be presented leading to the optimal design resulting in the lowest possible lithium contamination in the core.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

Comparison of Fuel Cycles for Lead-Lithium and Pure Lithium Liquid Metal Walls in a Magnetized Target Fusion Power Plant

General Fusion (GF) is developing an adaptable, commercial fusion power plant based on magnetized target fusion (MTF). The GF approach involves forming a spherical torus of deuterium-tritium plasma in a large (~4 m diameter) cavity formed in liquid metal, and then collapsing that cavity with an array of pneumatic piston drivers. The liquid metal is constantly flowing through the fusion chamber and out to processing systems where tritium and heat will be extracted using tritium extraction technologies and heat exchangers, respectively. Here, this study focuses on two candidate designs for the liquid metal blanket and first wall material for the General Fusion Magnetized Target Fusion (GF MTF) power plant and assesses their impact on the tritium fuel cycle. The first candidate is the lead lithium eutectic (LLE) and the second candidate is pure lithium (Li). It was found that the main differences between LLE and Li designs are the extraction technologies required to remove tritium from the blanket and the amount of tritium and its distribution within the facility. More than 80% of the in-process tritium inventory for the LLE design is contained in the isotope separation system, while for the Li design, over 60% of the in-process tritium inventory is contained within the blanket material. This is due to significant tritium retention by Li. For the Li blanket, the burden of tritium processing rests on the blanket extraction technology rather than the traditional exhaust processing route. Thus, the blanket extraction technology is a main driver of tritium inventory in the Li system and determines the subsequent interface with the tritium processing plant.

General Fusion↗
NASA NTRS1973

Gettering capsule for removing oxygen from liquid lithium systems

Capsule consisting of tantalum shell lined with tantalum screen and partially filled with lithium and pieces of yttrium is immersed in hot lithium stream. Oxygen is removed from stream by being absorbed by gettering capsule. Oxygen passes through capsule wall and into lithium inside capsule where it reacts with yttrium to form Y2O3.

Tower, L. K.↗
DOE OSTI2025

Active Divertor Heat Flux Control using Impurity Powder Dropper

Divertor plasma-facing components (PFCs) in a tokamak are typically designed to withstand average steady-state heat loads of about 5–10 MW/m 2 , a limit that applies to both solid and liquid lithium (LL) PFCs. Exceeding these design values can result in surface damage to tungsten PFCs or excessive lithium (Li) evaporation in liquid lithium divertor (LLD) PFCs. Since exceeding the divertor heat load limits has serious consequences, it is therefore prudent to develop a tool to reduce the divertor heat load and bring the heat load to within the design limit without affecting the plasma performance. Active low Z impurity injection such as Li has been suggested as a potential solution to mitigate excess heat flux as suggested previously, given that non-coronal radiation can be quite large ~ 20–30 MJ per mole of injected Li. Li is considered desirable for reducing the edge neutral recycling helping to improve plasma energy confinement. In this paper, we model the Impurity Power Dropper (IPD) to investigate its potential of divertor heat flux control. The IPD is typically located at the top of the tokamak device and uses a vertical drift tube of a few meters. In the 2 m drift tube case, the IPD powder is accelerated to ~ 6 m/sec before reaching the plasma with the upper divertor configuration, matching the condition for the in-board side pellet injection case. By modeling the IPD geometry we determined the IPD powder deposition profile, and thus the non-coronal radiation and ionization profiles in time as well. From the enhanced radiation power loss, it is therefore possible to reduce the divertor heat load using the divertor simulation code. In conclusion, the IPD divertor heat flux control can be tested in the facilities with IPD including ST-40, DIII-D, EAST, WEST and NSTX-U.

Active lithium injection↗
NASA NTRS2014

Developing the Pulsed Fission-Fusion (PuFF) Engine

In September 2013 the NASA Innovative Advanced Concept (NIAC) organization awarded a phase I contract to the PuFF team. Our phase 1 proposal researched a pulsed fission-fusion propulsion system that compressed a target of deuterium (D) and tritium (T) as a mixture in a column, surrounded concentrically by Uranium. The target is surrounded by liquid lithium. A high power current would flow down the liquid lithium and the resulting Lorentz force would compress the column by roughly a factor of 10. The compressed column would reach criticality and a combination of fission and fusion reactions would occur. Our Phase I results, summarized herein, review our estimates of engine and vehicle performance, our work to date to model the fission-fusion reaction, and our initial efforts in experimental analysis.

Adams, Robert B.↗
NASA NTRS2018

Pulsed Fission-Fusion (PuFF)

In September 2013 the NASA Innovative Advanced Concept (NIAC) organization awarded a phase I contract to the PuFF team. Our phase 1 proposal discussed a pulsed fission-fusion propulsion system that injected gaseous deuterium (D) and tritium (T) as a mixture in a column, surrounded concentrically by gaseous uranium fluoride (UF6) and then an outer shell of liquid lithium. A high power current would flow down the liquid lithium and the resulting Lorentz force would compress the column by roughly a factor of 10. The compressed column would reach criticality and a combination of fission and fusion reactions would occur. The fission reactions would further energize the fusion center, and the fusion reactions would generate neutrons that promote more complete burnup of the fission fuel. The lithium liner provides some help as a neutron reflector but also acts as a propulsive medium, being converted to plasma which is then expanded against a magnetic nozzle for thrust. The expansion of the (primarily) lithium plasma against the nozzle's magnetic field inducts a current that is used to charge the system for the next pulse. Our concept also included secondary injection of a Field Reversed Configuration (FRC) plasmoid that would provide a secondary compression direction, axially against the column, and push the column away from the injection manifold, increasing the manifold's survivability.Our phase 1 proposal included modeling the above process first under steady state assumptions and second under a time variant integration. We proposed including these results into a Mars concept vehicle and finally proposing promising conditions to be evaluated experimentally in Phase II. In phase I we quickly realized that we needed to modify our approach. Our steady state work was completed as proposed, and the results indicated that one, a two stage compression system was not needed and two, that we wanted to move away from UF6. The steady state model shows much more margin than expected, to the point that we may well reach breakeven with the Charger – 1 facility, a 572 kJ Marx bank currently under refurbishment at UAH. Additionally we found that using gaseous D-T and UF6, provided a relatively simple prospect of using a pulsed injector, made reaching criticality more difficult. The introduction of large amounts of fluorine meant a radiative sink, sapping power from the fusion plasma and was harder to handle. Therefore we moved to a solid uranium target that held D-T under pressure. In so doing we could move our target closer to criticality and remove any material that did not sustain the reaction.

Fusion↗
DOE OSTI2025

Lithium Divertor Targets and Walls for the ASTER Liquid Stellarator Reactor, Distributed Divertor

Stellarators may have advantages for certain liquid metal options as Plasma Facing Components (PFC) for divertor targets and first walls due to the wide range of possible magnetic configurations, which additionally are free of disruptions and fast field variations. In a previous work (V. Queral et al., IEEE Trans. Plasma Sci. 52, 2024), a concept of stellarator reactor (ASTER-CP) based on swirling Li-molten salts and liquid lithium floating on the molten salt as PFC was presented. The divertor matters were not studied then and, thus, they are being studied and experimentally tested now. The ASTER-CP reactor concept, the initial liquid metal experiments and potential concepts for the ASTER-CP divertor and first wall are reported. Concerning the experiments, several small scale experiments of galinstan in a small rotating cylinder under magnetic field have been produced, including one experiment with high viscosity galinstan-mixture for increased thickness of layer. An experiment of floating lithium on the molten salt LiCl-PbCl2 gave fast volatilization/decomposition of the molten salt. Particularly for divertors, the traditional free-flow, Capillary Porous Systems and ‘divertorlets’ have been studied for application to ASTER-CP. Surface waves (hot spots), lack of enough surface fluid turbulence and excessive fluid speed are the main issues found in fast free-flow. The perhaps original concept of Distributed Divertor and Equi-power Surface is tentatively proposed and studied, taking advantage of stellarator fields and low recycling regime.

distributed divertor↗
NASA NTRS1991

An investigation of voids formation mechanisms and their effects on freeze and thaw processes of lithium and lithium fluoride

The mechanisms of void formation during the cooldown and freezing of lithium coolant within the primary loop of SP-100 type systems are investigated. These mechanisms are: (1) homogeneous nucleation; (2) heterogeneous nucleation; (3) normal segregation of helium gas dissolved in liquid lithium; and (4) shrinkage of lithium during freezing. To evaluate the void formation potential due to segregation, a numerical scheme that couples the freezing and mass diffusion processes in both the solid and liquid regions is developed. The results indicated that the formation of He bubbles is unlikely by either homogeneous or heterogeneous nucleation during the cooldown process. However, homogeneous nucleation of He bubbles following the segregation of dissolved He in liquid lithium ahead of the solid-liquid interface is likely to occur. Results also show that total volume of He void is insignificant when compared to that of shrinkage voids. In viewing this, the subsequent research focuses on the effects of shrinkage void forming during freezing of lithium on subsequent thaw processes are investigated using a numerical scheme that is based on a single (solid/liquid) cell approach. The cases of lithium-fluoride are also investigated to show the effect of larger volume shrinkage upon freezing on the freeze and thaw processes. Results show that a void forming at the wall appreciably reduces the solid-liquid interface velocity, during both freeze and thaw, and causes a substantial rise in the wall temperature during thaw. However, in the case of Li, the maximum wall temperature was much lower than the melting temperature of PWC-11, which is used as the structure material in the SP-100 system. Hence, it is included that a formation of hot spots is unlikely during the startup or restart of the SP-100 system.

El-Genk, Mohamed S.↗
NASA NTRS1991

Mechanisms of voids formation during cooldown and freezing of lithium in SP-100 type systems

The mechanisms of void formation during the cooldown and freezing of lithium coolant within the primary loop of SP-100 type systems are investigated. These mechanisms are: (a) homogeneous nucleation, (b) heterogeneous nucleation, (c) normal segregation of helium gas dissolved in liquid lithium, and (d) shrinkage of lithium during freezing. To evaluate the void formation potential due to segregation, a numerical scheme that couples the freezing and mass diffusion processes in both the solid and liquid regions is developed. The results indicated that the formation of He bubbles is unlikely by either homogeneous or heterogeneous nucleation during the cooldown process. However, homogeneous nucleation of He bubbles following the segregation of dissolved He in liquid Lithium ahead of the solid-liquid interface is likely to occur. Results also show that total volume of He void is insignificant when compared to that of shrinkage voids.

Yang, Jae Y.↗
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↗
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.↗