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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 OSTI2015

High performance discharges in the Lithium Tokamak eXperiment with liquid lithium walls

The first-ever successful operation of a tokamak with a large area (40% of the total plasma surface area) liquid lithium wall has been achieved in the Lithium Tokamak eXperiment (LTX). These results were obtained with a new, electron beam-based lithium evaporation system, which can deposit a lithium coating on the limiting wall of LTX in a five-minute period. Preliminary analyses of diamagnetic and other data for discharges operated with a liquid lithium wall indicate that confinement times increased by 10× compared to discharges with helium-dispersed solid lithium coatings. Ohmic energy confinement times with fresh lithium walls, solid and liquid, exceed several relevant empirical scaling expressions. Spectroscopic analysis of the discharges indicates that oxygen levels in the discharges limited on liquid lithium walls were significantly reduced compared to discharges limited on solid lithium walls. Tokamak operations with a full liquid lithium wall (85% of the total plasma surface area) have recently started.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2020

Overview of lithium injection and flowing liquid lithium results from the US–China collaboration on EAST

A multi-institutional team consisting of national labs and universities has been conducting research to understand and control the plasma-material interface (PMI) to improve long pulse discharge control and performance in the EAST device in Hefei, China. One particular focus of the group has been to evaluate the performance of different plasma-facing component (PFC) technologies and materials, particularly lithium. Flowing liquid lithium PFC technologies are currently into their third generation of development and deployment in EAST, demonstrating recycling control and mitigation of edge-localized modes (ELMs). A lithium powder dropper has also been used to control ELMs while a lithium granule injector has demonstrated ELM triggering and pacing. To support this work, the facilities at the Center for Plasma Material Interactions such as HIDRA are being commissioned and utilized to develop/test the technologies before deployment on EAST. This paper will present a summary of the collaborative lithium PMI program on EAST.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2014

VUV/XUV measurements of impurity emission in plasmas with liquid lithium surfaces on LTX [VUV/XUV measurements of low recycling plasmas with liquid lithium surfaces on LTX]

The VUV/XUV spectrum has been measured on the Lithium Tokamak eXperiment (LTX) using a transmission grating imaging spectrometer (TGIS) coupled to a direct-detection x-ray charge-coupled device camera. TGIS data show significant changes in the ratios between the lithium and oxygen impurity line emission during discharges with varying lithium wall conditions. Lithium coatings that have been passivated by lengthy exposure to significant levels of impurities contribute to a large O/Li ratio measured during LTX plasma discharges. Furthermore, previous results have indicated that a passivated lithium film on the plasma facing components will function as a stronger impurity source when in the form of a hot liquid layer compared to a solid lithium layer. However, recent TGIS measurements of plasma discharges in LTX with hot stainless steel boundary shells and a fresh liquid lithium coating show lower O/Li impurity line ratios when compared to discharges with a solid lithium film on cool shells. In conclusion, these new measurements help elucidate the somewhat contradictory results of the effects of solid and liquid lithium on plasma confinement observed in previous experiments.

36 MATERIALS SCIENCE↗
DOE OSTI2015

Rock-salt structure lithium deuteride formation in liquid lithium with high-concentrations of deuterium: a first-principles molecular dynamics study

Because of lithium's possible use as a first wall material in a fusion reactor, a fundamental understanding of the interactions between liquid lithium (Li) and deuterium (D) is important. Here, we predict structural and dynamical properties of liquid Li samples with high concentrations of D, as derived from first-principles molecular dynamics simulations. Liquid Li samples with four concentrations of inserted D atoms (LiD$_{\beta}$ , $\beta =0.25$ , 0.50, 0.75, and 1.00) are studied at temperatures ranging from 470 to 1143 K. Densities, diffusivities, pair distribution functions, bond angle distribution functions, geometries, and charge transfer between Li and D atoms are calculated and analyzed. The analysis suggests liquid–solid phase transitions can occur at some concentrations and temperatures, forming rock-salt LiD within liquid Li. Finally, we observed the formation of some D 2 molecules at high D concentrations.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2020

Lithium splashing from flowing liquid lithium limiter and its effect on high confinement plasma performance in EAST tokamak

To study the potential applications of liquid lithium (Li) as plasma facing material, the third generation of flowing liquid Li (FLiLi) limiter has been designed and successfully tested in EAST tokamak H-mode discharges in the 2018 run campaign. Here, the stability of liquid Li on FLiLi surface, and Li splashing and its effect on the plasma performance are investigated in detail. Li splashing was observed in L-mode and H-mode discharges due to Rayleigh-Taylor (R-T) instability, resulting from electromagnetic forces J → x B → . Li splashing in low-parameter plasma decreased both core and edge electron temperatures, resulting in the degradation of plasma energy confinement and even H-L back transition, however, it almost did not influence plasma confinement performance in high-parameter plasma. These findings indicate that high-parameter plasma has higher tolerance to Li impurities than low-parameter plasma. By contrast, tungsten splashing from the melted layer on the upper divertor was also induced by the R-T instability but generally resulted in plasma disruption, which suggests lower tolerance to tungsten impurities in EAST tokamak.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
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 OSTI2013

High-flux neutron source based on a liquid-lithium target

A prototype compact Liquid Lithium Target (LiLiT), able to constitute an accelerator-based intense neutron source, was built. The neutron source is intended for nuclear astrophysical research, boron neutron capture therapy (BNCT) in hospitals and material studies for fusion reactors. The LiLiT setup is presently being commissioned at Soreq Nuclear research Center (SNRC). The lithium target will produce neutrons through the {sup 7}Li(p,n){sup 7}Be reaction and it will overcome the major problem of removing the thermal power generated by a high-intensity proton beam, necessary for intense neutron flux for the above applications. The liquid-lithium loop of LiLiT is designed to generate a stable lithium jet at high velocity on a concave supporting wall with free surface toward the incident proton beam (up to 10 kW). During off-line tests, liquid lithium was flown through the loop and generated a stable jet at velocity higher than 5 m/s on the concave supporting wall. The target is now under extensive test program using a high-power electron-gun. Up to 2 kW electron beam was applied on the lithium flow at velocity of 4 m/s without any flow instabilities or excessive evaporation. High-intensity proton beam irradiation will take place at SARAF (Soreq Applied Research Accelerator Facility) superconducting linear accelerator currently in commissioning at SNRC.

43 PARTICLE ACCELERATORS↗
DOE OSTI2012

Recent Progress in the NSTX/NSTX-U Lithium Program and Prospects for Reactor-Relevant Liquid-Lithium Based Divertor Development

Developing a reactor compatible divertor has been identified as a particularly challenging technology problem for magnetic confinement fusion. While tungsten has been identified as the most attractive solid divertor material, the NSTX/NSTX-U lithium (Li) program is investigating the viability of liquid lithium (LL) as a potential reactor compatible divertor plasma facing component (PFC) . In the near term, operation in NSTX-U is projected to provide reactor-like divertor heat loads < 40 MW/m^2 for 5 s. During the most recent NSTX campaign, ~ 0.85 kg of Li was evaporated onto the NSTX PFCs where a ~50% reduction in heat load on the Liquid Lithium Divertor (LLD) was observed, attributable to enhanced divertor bolometric radiation. This reduced divertor heat flux through radiation observed in the NSTX LLD experiment is consistent with the results from other lithium experiments and calculations. These results motivate an LL-based closed radiative divertor concept proposed here for NSTX-U and fusion reactors. With an LL coating, the Li is evaporated from the divertor strike point surface due to the intense heat. The evaporated Li is readily ionized by the plasma due to its low ionization energies, and the ionized Li ions can radiate strongly, resulting in a significant reduction in the divertor heat flux. Due to the rapid plasma transport in divertor plasma, the radiation values can be significantly enhanced up to ~ 11 MJ/cc of LL. This radiative process has the desired function of spreading the focused divertor heat load to the entire divertor chamber facilitating the divertor heat removal. The LL divertor surface can also provide a "sacrificial" surface to protect the substrate solid material from transient high heat flux such as the ones caused by the ELMs. The closed radiative LLD concept has the advantages of providing some degree of partition in terms of plasma disruption forces on the LL, Li particle divertor retention, and strong divertor pumping action from the Li-coated divertor chamber wall. By operating at a lower temperature than the first wall, the LLD can serve to purify the entire reactor chamber, as impurities generally migrate toward lower temperature Li-condensed surfaces. To maintain the LL purity, a closed LL loop system with a modest capacity (e.g., ~ 1 Liter/sec for ~ 1% level "impurities") is envisioned for a steady-state 1 GW-electric class fusion power plant.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2012

Comparison of H-Mode Plasmas Diverted to Solid and Liquid Lithium Surfaces

Experiments were conducted with a Liquid Lithium Divertor (LLD) in NSTX. Among the goals was to use lithium recoating to sustain deuterium (D) retention by a static liquid lithium surface, approximating the ability of flowing liquid lithium to maintain chemical reactivity. Lithium evaporators were used to deposit lithium on the LLD surface. Improvements in plasma edge conditions were similar to those with lithiated graphite plasma-facing components (PFCs), including an increase in confinement over discharges without lithiumcoated PFCs and ELM reduction during H-modes. With the outer strike point on the LLD, the D retention in the LLD was about the same as that for solid lithium coatings on graphite, or about two times that achieved without lithium PFC coatings. There were also indications of contamination of the LLD surface, possibly due erosion and redeposition of carbon from PFCs. Flowing lithium may thus be needed for chemically active PFCs during long-pulse operation.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2013

High-power liquid-lithium jet target for neutron production

A compact liquid-lithium target (LiLiT) was built and tested with a high-power electron gun at the Soreq Nuclear Research Center. The lithium target, to be bombarded by the high-intensity proton beam of the Soreq Applied Research Accelerator Facility (SARAF), will constitute an intense source of neutrons produced by the {sup 7}Li(p,n){sup 7}Be reaction for nuclear astrophysics research and as a pilot setup for accelerator-based Boron Neutron Capture Therapy. The liquid-lithium jet target acts both as neutron-producing target and beam dump by removing the beam thermal power (>5 kW, >1 MW/cm{sup 3}) with fast transport. The target was designed based on a thermal model, accompanied by a detailed calculation of the {sup 7}Li(p,n) neutron yield, energy distribution, and angular distribution. Liquid lithium is circulated through the target loop at ∼200 °C and generates a stable 1.5 mm-thick film flowing at a velocity up to 7 m/s onto a concave supporting wall. Electron beam irradiation demonstrated that the liquid-lithium target can dissipate electron power areal densities of >4 kW/cm{sup 2} and volume power density of ∼2 MW/cm{sup 3} at a lithium flow of ∼4 m/s while maintaining stable temperature and vacuum conditions. The LiLiT setup is presently in online commissioning stage for high-intensity proton beam irradiation (1.91–2.5 MeV, 1–2 mA) at SARAF.

43 PARTICLE ACCELERATORS↗
DOE OSTI2016

Upgrades toward high-heat flux, liquid lithium plasma-facing components in the NSTX-U

Liquid metal plasma-facing components (PFCs) provide numerous potential advantages over solid-material components. One critique of the approach is the relatively less developed technologies associated with deploying these components in a fusion plasma-experiment. Exploration of the temperature limits of liquid lithium PFCs in a tokamak divertor and the corresponding consequences on core operation are a high priority informing the possibilities for future liquid lithium PFCs. An all-metal NSTX-U is envisioned to make direct comparison between all high-Z wall operation and liquid lithium PFCs in a single device. By executing the all-metal upgrades incrementally, scientific productivity will be maintained while enabling physics and engineering-science studies to further develop the solid- and liquid-metal components. Six major elements of a flowing liquid-metal divertor system are described and a three-step program for implementing this system is laid out. The upgrade steps involve the first high-Z divertor target upgrade in NSTX-U, pre-filled liquid metal targets and finally, an integrated, flowing liquid metal divertor target. As a result, two example issues are described where the engineering and physics experiments are shown to be closely related in examining the prospects for future liquid metal PFCs.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2014

Electrical detection of liquid lithium leaks from pipe joints

A test stand for flowing liquid lithium is under construction at Princeton Plasma Physics Laboratory. As liquid lithium reacts with atmospheric gases and water, an electrical interlock system for detecting leaks and safely shutting down the apparatus has been constructed. A defense in depth strategy is taken to minimize the risk and impact of potential leaks. Each demountable joint is diagnosed with a cylindrical copper shell electrically isolated from the loop. By monitoring the electrical resistance between the pipe and the copper shell, a leak of (conductive) liquid lithium can be detected. Any resistance of less than 2 kΩ trips a relay, shutting off power to the heaters and pump. The system has been successfully tested with liquid gallium as a surrogate liquid metal. The circuit features an extensible number of channels to allow for future expansion of the loop. To ease diagnosis of faults, the status of each channel is shown with an analog front panel LED, and monitored and logged digitally by LabVIEW.

42 ENGINEERING↗
DOE OSTI2015

Investigation of parameters of interaction of hydrogen isotopes with liquid lithium and lithium capillary-porous system under reactor irradiation

In this study, the effect of reactor irradiation on the processes of interaction of hydrogen with liquid lithium and a lithium capillary-porous system (CPS) is considered. The experiments are carried out by the gas-absorption method with use of a specially designed ampoule device. The results of investigation of the interaction of hydrogen with liquid lithium and a lithium CPS under conditions of reactor irradiation are described; namely, these are the temperature dependences of the rate constant for the interaction of hydrogen with liquid lithium at different reactor powers, the activation energies of the processes, and the pre-exponential factor in the Arrhenius dependence. The effect of increasing absorption of hydrogen by the samples under investigation as a result of the reactor irradiation is fixed. The effect can be explained by increasing mobility of hydrogen in liquid lithium due to hot spots in lithium bulk and the interaction of helium and tritium ions (formed as a result of the nuclear reaction of {sup 6}Li with neutron) with a surface hydride film.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Wetting of liquid lithium on fusion-relevant materials microtextured by femtosecond laser exposure

As the use of liquid metals in plasma facing components becomes more widespread, it is important to investigate how these liquid metals interact with the surfaces onto which they are deposited. An important example of these interactions is the ability to control liquid metal wettability on fusion relevant substrates. Here in this work, we explore the influence of femtosecond laser induced nanostructured surfaces on the wetting degree of liquid lithium versus temperature. Three material candidates as a lithium wall in magnetic fusion devices have been investigated: molybdenum, tungsten and 304 L stainless steel. Laser parameters were tuned to induce periodical self-organized nanostructures (ripples or LIPSS) formation on each material. Wettability of laser treated materials was changed from lithium-philic to lithium-phobic for temperatures beyond 320 °C - 360 °C compared to untreated material. The effect of both laser induced topography and chemistry are quantified to explain the observed liquid lithium contact angles on each material. Finally, it was shown that topography in the form of self-organized periodical nanostructures as well as the surface chemistry in the form of oxides enrichment, both induced by a single step laser process, strongly influence the wetting degree of liquid lithium and enhance lithium-phobicity at high temperatures.

36 MATERIALS SCIENCE↗
DOE OSTI2016

Conceptual design of a pre-loaded liquid lithium divertor target for NSTX-U

In this study, a conceptual design for a pre-filled liquid lithium divertor target for the National Spherical Torus Experiment Upgrade (NSTX-U) is presented. The design is aimed at facilitating experiments with high lithium flux from the plasma facing components (PFCs) in NSTX-U and investigating the potential of capillary based liquid lithium components. In the design, lithium is supplied from a reservoir in the PFC to the plasma facing surface via capillary action in a wicking structure. This working principle is also demonstrated experimentally. Next, a titanium zirconium molybdenum (TZM) prototype design is presented, required to withstand a steady state heat flux peaking at 10 MW m –2 for 5 s and edge localized modes depositing (130 kJ in 2 ms at 10 Hz). The main challenge is to sufficiently reduce the thermal stresses. This is achieved by dividing the surface into brushes and filling the slots in between with liquid lithium. The principle of using this liquid “interlayer” allows for thermal expansion and simultaneously heat conduction, and could be used to significantly reduce the demands to solids in future PFCs. Lithium flow to the surface is analyzed using a novel analytical model, ideally suited for design purposes. Thermal stresses in the PFC are analyzed using the finite element method. As a result, the requirements are met, and thus a prototype will be manufactured for physical testing.

42 ENGINEERING↗
DOE OSTI2021

Structured large-pore foams improve thermal performance of LiMIT-style liquid lithium PFC

As magnetically confined fusion devices improve, the conditions at the walls become increasingly intense. Plasma facing components (PFCs) must withstand these extreme heat and particle loads without damage or degradation. Liquid lithium PFCs are known to be quite resilient, and the presence of lithium also serves to improve plasma properties. The liquid metal infused trench (LiMIT) concept is an open surface liquid lithium PFC design that has been tested extensively at the University of Illinois and in fusion devices around the world. LiMIT utilizes thermoelectric magnetohydrodynamics (TEMHD) to passively drive liquid lithium flow. This work demonstrates an extension of the LiMIT trench geometry to three dimensions. Additively manufactured large pore metallic foams maintain TEMHD drive while drastically improving heat flux handling and resistance to lithium dryout, a phenomenon where locally high TEMHD forces depresses the lithium level and exposes underlying solid structure. COMSOL multiphysics modeling of the system yields insight into the forces at play in dryout development, and shows the 3D structures can eliminate dryout. Here, low heat proof-of-concept experimental testing of the system matches computational results, and high heat flux electron beam tests more than double the proven operational range of a LiMIT-style PFC, to 6.8 MW m –2 , with no indications of dryout or impending damage.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Design and Analysis of Liquid Lithium Plasma Facing Components

In this study, liquid metal (LM) on plasma-facing components (PFCs) creates a renewable protective cover enhancing power exhaust and protecting the solid substrate. Additionally, energy confinement improvement through particle pumping can be achieved in the case of liquid lithium (LL). A new variant of PFCs was recently introduced at Princeton Plasma Physics Laboratory (PPPL) where a porous wall is used to stabilize the LM surface, while using magnetohydrodynamic (MHD) drive to push the LM flow inside the component. This arrangement allows efficient heat exhaust, and enhanced control of the LM surface temperature, leading to spatial control of evaporation and condensation of LL on the plasma interface. This feature is particularly attractive when vapor shielding is introduced to allow heat flux redistribution. This system has the advantage that as the heat flux increases, the evaporation rate will increase, while decreasing heat flux will decrease the evaporation rate of lithium, ideally creating a feedback effect which could self-regulate the amount of lithium evaporated. Analytical and numerical models for LL PFCs were developed in PPPL. To calculate the target temperature distribution for the case of evaporation from the divertor, we apply an iterative process allowing two-way coupling between the fluid-kinetic analysis of plasma using SOLPS-ITER code and the flow and heat transfer analysis of the PFC using an analytical model. In the last stage, the results are validated using computational fluid dynamics (CFD) analysis with customized version of the CFX code from ANSYS. CFX was modified at PPPL to allow MHD analysis at the high magnetic field typical for fusion applications. Results of the analysis for NSTX-U tokamak conditions will be presented.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗