Liquid Lithium Divertor Research Articles

Design of a 3D-printed liquid lithium divertor target plate and its interaction with high-density plasma

A liquid Li divertor is a promising alternative for future fusion devices. In this work a new divertor model is proposed, which is processed by 3D-printing technology to accurately control the size of the internal capillary structure. At a steady-state heat load of 10 MW m−2, the thermal stress of the tungsten target is within the bearing range of tungsten by finite-element simulation. In order to evaluate the wicking ability of the capillary structure, the wicking process at 600 °C was simulated by FLUENT. The result was identical to that of the corresponding experiments. Within 1 s, liquid lithium was wicked to the target surface by the capillary structure of the target and quickly spread on the target surface. During the wicking process, the average wicking mass rate of lithium should reach 0.062 g s−1, which could even supplement the evaporation requirement of liquid lithium under an environment > 950 °C. Irradiation experiments under different plasma discharge currents were carried out in a linear plasma device (SCU-PSI), and the evolution of the vapor cloud during plasma irradiation was analyzed. It was found that the target temperature tends to plateau despite the gradually increased input current, indicating that the vapor shielding effect is gradually enhanced. The irradiation experiment also confirmed that the 3D-printed tungsten structure has better heat consumption performance than a tungsten mesh structure or multichannel structure. These results reveal the application potential and feasibility of a 3D-printed porous capillary structure in plasma-facing components and provide a reference for further liquid−solid combined target designs.

Conceptual design of a 3D printed liquid lithium divertor test modular for EAST

The utilization of Liquid Metal Divertor (LMD) stands as a promising prospect for the divertor component in DEMO. To explore the vapor shielding effect and provide technical accumulation for the development of DEMO LMD, the imperative development of a high-performance LMD test module has arisen. The Experimental Advanced Superconducting Tokamak (EAST) has amassed substantial expertise in LM first wall design and experimentation, rendering it suitably equipped for experimental pursuits. Therefore, this paper aims to design a test module running in the mid-plane of EAST to accumulate technical experience for the application of 3D printed capillary porous system (CPS) in LMD. Drawing upon prior experience, specific design requirements are articulated. Lithium is chosen as the operational LM due to its favorable balance between heat dissipation capacity and compatibility with the core plasma. The conceptual design encompasses active water cooling components, a CPS, and a lithium channel. Thermal analysis is conducted to ascertain the prevention of excessive evaporation or frozen during operation. The maximum power handling capacity of this design is projected, with practical solutions outlined for inter-component thermal management. The results affirm the viability of this design within the thermal load parameters of the EAST experiment, offering invaluable insights for designers exploring diverse LMD configurations.

Simulations of lithium impurity transport with a liquid lithium divertor using EMC3-EIRENE and ITCD

Simulations of lithium (Li) impurity transport behaviour with a liquid lithium divertor on the EAST tokamak have been performed using the fluid code EMC3-EIRENE and the Monte Carlo code ITCD. The interactions between particles and the wall surface, like the gyration-induced scrape-off effect, sheath acceleration, kinetic self-sputtering and reflection, are typically ignored or simplified in fluid models. In order to uncover the impacts of the scrape-off effect, sheath acceleration, kinetic self-sputtering and reflection on Li distribution, detailed analysis of the density distributions of Li1+ and Li2+ ions at the divertor region have been carried out by ITCD modelling. The newly-upgraded ITCD code can principally reproduce the distribution patterns of Li ions as well as EMC3-EIRENE under identical plasma conditions on EAST. However, due to the scrape-off effect and sheath acceleration, the densities of Li1+ and Li2+ ions simulated by ITCD are evidently reduced compared to the EMC3-EIRENE simulations. The kinetic self-sputtering by redeposited Li particles in ITCD can increase the densities of the Li ions. Moreover, the kinetic reflection database calculated by SRIM code has been employed by ITCD, which leads to an increase in the Li densities compared to the reflection approach used in EMC3-EIRENE. Overall, by adding the aforesaid physical effects in ITCD, the total number of Li ions shows an obvious reduction compared to the EMC3-EIRENE modelling.

Tolerable Stationary Heat Loads to Liquid Lithium Divertor Targets

An 0D model is proposed that makes it possible to estimate the limiting stationary heat loads to the targets covered with liquid lithium (LL) layer, taking into account the effects of vapor shielding by sputtered and evaporated LL and hydrogen recycling. Several models of cooled target substrates are considered in which the LL layer facing the plasma is placed. For the considered substrate models, a parametric analysis of the tolerable stationary heat loads to the target on the substrate thickness, the effective cooling energy per particle of sputtered lithium, and the lithium prompt redeposition factor was carried out. It is shown that, at a small substrate thickness, the choice of the substrate model has a significant impact on the tolerable heat loads. It is also shown that even at unrealistically large values of the effective cooling energy, the dissipation of lithium remains modest. This means that in regimes with a high power coming from the core plasma to the edge, the injection of an additional radiator is required. Finally, it is shown that one of the most effective ways to increase the tolerable stationary heat loads would be to reduce the thickness of the target substrate.

Simulation of lithium flow, redeposition, and vapor shielding in liquid lithium divertor of T-15MD tokamak with SOLPS 4.3 code

The erosion module of the SOLPS 4.3 code package for the liquid Li divertor target is developed to include the thermal sputtering, evaporation, and prompt redeposition effects. Eroded Li atoms are treated within a fully kinetic approach. Simulations for the T-15MD tokamak with liquid Li divertor targets are performed for various operational regimes. The main parameters characterizing the divertor operation are identified. The Li shielding effect is found significant at the input power P SOL > 15 MW. The target conditions remain reasonable in terms of both the erosion rate and target temperature but core plasma dilution with Li remains a problem. The Li flow patterns in the scrape-of-layer (SOL) are analyzed in detail. About 80% of eroded Li is redeposited back on the target. The energy dissipated per an eroded Li particle, E sh, is calculated and its values are in the range of 100–200 eV. These values are 2–4 times lower than those obtained from a simple 0D estimation of E sh. Therefore, relying on simple 0D models for the analysis of Li divertor performance may cause an overestimation of the shielding efficiency.

The development of the MHD film flow numerical simulation platform

<p indent="0mm">In this paper, magnetic confinement fusion is introduced focusing on its fluid mechanics relevant to a liquid lithium divertor. The main research problem is the flow stability of a liquid metal film flow under the influence of a magnetic field. Therefore, a linear stability analysis numerical calculation platform and a direct numerical simulation platform are developed to understand the effect of magnetohydrodynamics on the flow stability of thin films. These platforms will be used to reveal the physical mechanism of a magnetic field on the film flow stability and to guide the design and implementation of a liquid lithium divertor.

Lithium, a path to make fusion energy affordable

In this tutorial article, we review the technological, physics, and economic basis for a magnetic fusion device utilizing a flowing liquid lithium divertor (molten metal velocity in the range of cm/s) and operating in a low-recycling plasma regime. When extrapolated to magnetic fusion reactor scale, the observed effects of a liquid lithium boundary on recycling reduction, confinement increase, and anomalous heat transport mitigation may offer a fundamentally distinct and promising alternative route to fusion energy production. In addition, this lithium-driven low recycling regime could accelerate fusion's commercial viability since such a device would be smaller, dramatically decreasing plant and electricity costs if all technological complexities are solved. First, the theoretical basis of the energy confinement and fusion performance as well as the related possibilities of low recycling regimes driven by flowing lithium plasma-facing components are reviewed. Then the paper emphasizes the technological obstacles that need to be overcome for developing the necessary systems for such a flowing liquid lithium solution at reactor scale and details how many of these have been overcome at laboratory and/or proof-of-concept scale. Finally, the current and planned scientific and engineering endeavors being performed at the University of Illinois at Urbana-Champaign regarding this alternative reactor option are discussed.

Modeling the vapor shielding of a liquid lithium divertor target using SOLPS 4.3 code

In this letter, we report the very first results of SOLPS4.3 simulations of a liquid lithium (Li) divertor, including vapor shielding effects. A peculiarity of Li as a target coating material is the strong dependence of the erosion rate on the target temperature. We have implemented a new erosion model in SOLPS, taking this dependence into account. Simulations of the T15-MD tokamak divertor with Li-coated divertor targets have been performed. Li erosion is determined by physical sputtering, evaporation, and thermal sputtering. The results of the simulations show that a shielding effect occurs, providing a reduction of the target heat flux to values below 8 MW m−2. At the same time, the upstream plasma dilution in the high-power regimes, where shielding is most efficient, is very strong, meaning that in practical terms, it may be difficult to rely on the shielding effect for heat flux control in the configuration considered.

Influence of Krypton Seeding on EU DEMO Operation with Lithium Divertor

DEMO reactor with liquid lithium divertor operation in presence of krypton seeding is analyzed. Integrated core-scrape-off layer-divertor code COREDIV is used. Sputtering model is revised and sputtering dependence on the divertor surface temperature fitted to the experimental data is added. It was found that Kr seeding is beneficial for DEMO operation. Plasma fuel dilution by lithium ions is reduced to the level of 1% for high level of Kr seeding and fusion power is comparable to the modelling results for argon or neon seeded DEMO with tungsten divertor. Simulations show that high confinement mode is possible in presence of Kr seeding (power across the separatrix stays above the L-H threshold).

Active Radiative Liquid Lithium Divertor for Handling Transient High Heat Flux Events

The extreme heat flux anticipated in fusion reactor divertor plasma facing components (PFCs) is perhaps the most challenging technology issue for fusion energy development. Most divertor PFCs are designed based on the maximum steady-state operational limits. Application of lithium (Li) in NSTX resulted in improved H-mode confinement, H-mode power threshold reduction, and reduction in the divertor peak heat flux while maintaining essentially Li-free core plasma operation even during H-modes. These promising Li results in NSTX and related modeling calculations motivated the passive and active radiative liquid lithium divertor concepts (RLLD and ARLLD) (Ono et al. in Nucl Fusion 53:113030, 2013; Fusion Eng Des 89:2838, 2014). This radiative process has the desired effect of spreading the localized divertor heat load to the rest of the divertor chamber wall surfaces, facilitating divertor heat removal with relatively small amount of lithium ~ few moles/s to handle the expected steady-state high divertor heat load. However, in addition to the high steady-state heat flux, the fusion reactor divertor PFCs could also experience significant transient heat flux such as ELMs and/or other magnetic reconnection events which can deposit large transient heat flux onto the divertor PFCs. If unprotected, it could damage the divertor PFC surfaces which could lead to a highly undesirable unplanned shutdown for PFC repair and/or replacement. In this paper, we explore feasibility of LLD and ARLLD concepts in handling such transient heat flux to protect the divertor PFCs from the extreme transient heat flux while maintaining the normal plasma operations. We also suggest a possible implementation technique using inductive pellet injector for the reactor PFC protection from transient heat flux which can be tested on NSTX-U.

Vapor shielding of liquid lithium divertor target during steady state and transient events

We develop a 0D model of vapor shielding including a vast variety of processes in the vapor cloud and plasma-surface interactions. The model is applied to a liquid lithium (Li) divertor with targets covered by a self-restoring Li coating. Calculations show that in the steady state, the target heat flux saturates with increasing external heat flux at approximately 7.5 MW m−2. This value may seem very optimistic. However, we show that a significant part of the external heat is removed with the eroded particles. Hence, the maximum sustainable heat load is restricted by the coating renewal rate. The response of the coating to pulsed loads similar to the ITER-scale type I edge localized modes (ELMs) demonstrates high shielding efficiency. Similar to the steady-state regime, the coating renewal rate is the critical quantity for the target survival under pulsed heat loads. Surviving an ELM with an external heat load of 100 MW m−2 requires a fast coating recovery rate (1026 m−2 s−1) or a substantial reservoir of Li on the surface, which may be a concern for some target designs.

Integrated power exhaust modelling for DEMO with lithium divertor

A DEMO reactor with a liquid lithium divertor setup is considered. The simulation is performed for EU DEMO 2018 configuration using COREDIV code which self-consistently solves core and scrape-off layer transport equations for plasma and impurity. Influence of the sputtering, prompt redeposition and evaporation of the liquid lithium divertor was taken into account. Two operation regimes were identified. In the first regime evaporation rate is suppressed by the cooling system and is of the order of the sputtering rate. Scrape-off layer plasma characterises with low radiation and high total power to the plates. In the second regime the evaporation rate is much higher than the sputtering rate. High Li influx reduces temperature in the divertor region. Lithium radiation is increased reducing power to the plates. Plasma parameters in the core are comparable for the two regimes and the fusion gain parameter Q drops only slightly. Both regimes characterise with high plasma dilution. Power across the separatrix stays above the L-H threshold.

Power handling and vapor shielding of pre-filled lithium divertor targets in Magnum-PSI

To develop realistic liquid lithium divertors for future fusion reactors, this paper aims to improve the understanding of their power handling capabilities. A liquid lithium divertor target prototype, designed to facilitate liquid metal experiments in tokamaks, was tested in Magnum-PSI. The target has an internal reservoir pre-filled with lithium and aims to passively re-supply the textured plasma facing surface during operation. To assess the power handling capability the target was exposed to helium plasmas with increasing power flux density in the linear plasma device Magnum-PSI. The temperature response of lithium targets was recorded via an infrared camera, and compared to finite element method modeling taking into account dissipation via lithium in the plasma. It was found that the target works as intended and can take up to 9 1 MW m−2 for 10 s before the mesh layer was damaged, and could continue operating at higher power densities even after being damaged. The total lifetime of the targets was up to 100 s. Overall the targets are found suitable for use in tokamak experiments. Additionally, a central surface temperature evolution indicative of vapor shielding was observed on intact targets. Predicting the target temperature (and consequently the evaporation rates and thermal stresses) is considered very relevant for the design of lithium divertor targets for DEMO. The observed temperature response could indeed be replicated through modeling, which showed that a significant power fraction was dissipated by the lithium in the plasma.

Power handling limit of liquid lithium divertor targets

A model is formulated to make a first estimate of the maximum tolerable power of liquid lithium (Li) divertor targets, and to gain insight into their behaviour in terms of Li loss rate and surface temperature. The model, formulated as a simple analytical expression, states that the incoming power is balanced by heat conduction through the target and by the Li which dissipates energy via evaporation, radiation and ion-neutral friction. A target is considered to fail when the net Li loss flux from the surface exceeds the available supply. The model is evaluated over a range of input parameters: Li supply rate, surface layer thickness, redeposition coefficient, and dissipated energy per Li particle lost to the plasma. Based on the results, first, surface temperature locking is expected above a deposited power of ∼10 MW m−2. Second, Li targets are expected to be extremely robust against power deposited during short transient events. A surface layer thickness of 50 micron is sufficient to withstand 60 MJ m−2 vertical displacement events or 20 MJ m−2 disruptions.

Liquid lithium loop system to solve challenging technology issues for fusion power plant

Steady-state fusion power plant designs present major divertor technology challenges, including high divertor heat flux both in steady-state and during transients. In addition to these concerns, there are the unresolved technology issues of long term dust accumulation and associated tritium inventory and safety issues. It has been suggested that radiation-based liquid lithium (LL) divertor concepts with a modest lithium-loop could provide a possible solution for these outstanding fusion reactor technology issues, while potentially improving reactor plasma performance. The application of lithium (Li) in NSTX resulted in improved H-mode confinement, H-mode power threshold reduction, and reduction in the divertor peak heat flux while maintaining essentially Li-free core plasma operation even during H-modes. These promising results in NSTX and related modeling calculations motivated the radiative liquid lithium divertor concept and its variant, the active liquid lithium divertor concept, taking advantage of the enhanced or non-coronal Li radiation in relatively poorly confined divertor plasmas. To maintain the LL purity in a 1 GW-electric class fusion power plant, a closed LL loop system with a modest circulating capacity of ~1 l s−1 is envisioned. We examined two key technology issues: (1) dust or solid particle removal and (2) real time recovery of tritium from LL while keeping the tritium inventory level to an acceptable level. By running the LL-loop continuously, it can carry the dust particles and impurities generated in the vacuum vessel to the outside where the dust/impurities can be removed by relatively simple dust filter, cold trap and/or centrifugal separation systems. With ~1 l s−1 LL flow, even a small 0.1% dust content by weight (or 0.5 g s−1) suggests that the LL-loop could carry away nearly 16 tons of dust per year. In a 1 GW-electric (or ~3 GW fusion power) fusion power plant, about 0.5 g s−1 of tritium is needed to maintain the fusion fuel cycle assuming ~1% fusion burn efficiency. It appears feasible to recover tritium (T) in real time from LL while maintaining an acceptable T inventory level. Laboratory tests are being conducted to investigate T recovery feasibility with the surface cold trap concept.

Hydrogen retention studies on lithiated tungsten exposed to glow discharge plasmas under varying lithiation environments using Thermal Desorption Spectroscopy and mass spectrometry

For the design of a Fusion Reactor based on a liquid lithium divertor target and a tungsten first wall at high temperature, the interaction of the wall material with plasmas of significant lithium content must be assessed, as issues like fuel retention, tungsten embrittlement and enhanced sputtering may represent a showstopper for the selection of the first wall material compatible with the presence of liquid metal divertor.In this work we address this topic for the first time at the laboratory level, hot W samples (100°C) have been exposed to Glow Discharges of H2 or Li-seeded H2 followed by in situ thermal desorption studies (TDS) of the uptake of H2 on the samples. Pure and pre-lithiated tungsten was investigated in order to evaluate the differential effect of Li ion implantation on H retention. Global particle balance was also used for the determination of trapped H into the full W wall of the plasma chamber. A factor of 3-4 lower retention was deduced for samples and main W wall exposed to H/Li plasma than that measured on pre-lithiated W.

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

In this work, 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 10MWm−2 for 5s and edge localized modes depositing (130kJ in 2ms at 10Hz). 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. The requirements are met, and thus a prototype will be manufactured for physical testing.

Liquid lithium applications for solving challenging fusion reactor issues and NSTX-U contributions

Steady-state fusion reactor operation presents major divertor technology challenges, including high divertor heat flux both steady-state and transients. In addition, there are unresolved issues of long term dust accumulation and associated tritium inventory and safety concerns (Federici et al., 2001) [1]. It has been suggested that radiative liquid lithium divertor concepts with a modest lithium-loop could provide a possible solution for these outstanding fusion reactor technology issues, while potentially improving reactor plasma performance (Ono et al., 2013, 2014) [2,3]. The application of lithium (Li) in NSTX resulted in improved H-mode confinement, H-mode power threshold reduction, and reduction in the divertor peak heat flux while maintaining essentially Li-free core plasma operation even during H-modes. These promising results in NSTX and related modeling calculations motivated the radiative liquid lithium (LL) divertor (RLLD) concept (Ono et al., 2013) [2] and its variant, the active liquid lithium divertor concept (ARLLD) (Ono et al., 2014) [3], taking advantage of the enhanced non-coronal Li radiation in relatively poorly confined divertor plasmas. It was estimated that only a few moles/s of lithium injection would be needed to significantly reduce the divertor heat flux in a tokamak fusion power plant. By operating at lower temperatures ≤450°C than the first wall ∼600–700°C, the LL-covered divertor chamber wall surfaces can serve as an effective particle pump, as impurities generally migrate toward lower temperature LL divertor surfaces. To maintain the LL purity, a closed LL loop system with a modest circulating capacity of ∼1l/s (l/s) is envisioned to sustain the steady-state operation of a 1GW-electric class fusion power plant. By running the Li loop continuously, it can carry the dust particles and impurities generated in the vacuum vessel to outside where the dust/impurities are removed by relatively simple filter and cold/hot trap systems. Using a cold trap system, it can recover tritium (T) in real time from LL at a rate of ∼0.5g/s needed to sustain the fusion reaction while minimizing the T inventory. With an expected T fraction of ≤0.7%, an acceptable T inventory level can be achieved. In NSTX-U (Menard et al., 2012; Ono et al., 2015) [4,5], preparations are now underway to elucidate the physics of Li plasma interactions with a number of Li application tools and Li radiation spectroscopic instruments. The NSTX-U Li evaporator, which provides Li coating over the lower divertor plate, can offer important information on the RLLD concept, and the Li granule injector will test some of the key physics issues for the ARLLD concept. A LL loop is also being prepared off-line for prototyping future use on NSTX-U.

Hydrogen retention in lithium on metallic walls from “in vacuo” analysis in LTX and implications for high-Z plasma-facing components in NSTX-U

The application of lithium to plasma-facing components (PFCs) has long been used as a technique for wall conditioning in magnetic confinement devices to improve plasma performance. Determining the characteristics of PFCs at the time of exposure to the plasma, however, is difficult because they can only be analyzed after venting the vacuum vessel and removing them at the end of an operational period. The Materials Analysis and Particle Probe (MAPP) addresses this problem by enabling PFC samples to be exposed to plasmas, and then withdrawn into an analysis chamber without breaking vacuum. The MAPP system was used to introduce samples that matched the metallic PFCs of the Lithium Tokamak Experiment (LTX). Lithium that was subsequently evaporated onto the walls also covered the MAPP samples, which were then subject to LTX discharges. In vacuo extraction and analysis of the samples indicated that lithium oxide formed on the PFCs, but improved plasma performance persisted in LTX. The reduced recycling this suggests is consistent with separate surface science experiments that demonstrated deuterium retention in the presence of lithium oxide films. Since oxygen decreases the thermal stability of the deuterium in the film, the release of deuterium was observed below the lithium deuteride dissociation temperature. This may explain what occurred when lithium was applied to the surface of the NSTX Liquid Lithium Divertor (LLD). The LLD had segments with individual heaters, and the deuterium-alpha emission was clearly lower in the cooler regions. The plan for NSTX-U is to replace the graphite tiles with high-Z PFCs, and apply lithium to their surfaces with lithium evaporation. Experiments with lithium coatings on such PFCs suggest that deuterium could still be retained if lithium compounds form, but limiting their surface temperatures may be necessary.

A physical model of an ejection suppressed CPS liquid lithium divertor target

A physical model has been developed which includes high temperature liquid lithium evaporation, the expanding motion of the liquid lithium vapour cloud, the shielding effects of the vapour cloud on incident plasma particle bombardments, ejection suppressed analysis and a perpendicular field proposal, and photon radiation, heat flux and transport in the lithium vapour cloud plasma. The engineering outline design scheme and the relevant parameters for the liquid lithium surface divertor target plate configured by discrete tiny capillary arrays have been established. Splashing can be suppressed by utilizing discrete and electrical insulating capillary porous systems (CPSs), since the conductivity among the capillary cells has been cut off by adopting a special kind of ceramic composite material made of a non-conducting and unbreakable composite which is able to withstand high temperatures. The formula to describe the temperature-dependent evaporation power has been derived. The maximum temperature increases of the discrete plasma-facing liquid lithium surface divertor target plate have been compared under the high energy flux deposition of 10 MJ m−2 during a 1 ms time duration with or without evaporation power. The results show that a high surface heat load can be withstood by the designed discrete plasma-facing liquid lithium surface divertor target plate due to violent evaporation. The energy deposition of incident energetic particles and weakly relativistic electrons from the scrape-off layer have been calculated. A laboratory experimental facility to simulate liquid lithium surface interactions with plasma has been set up. Research on lithium evaporation, re-deposition and ejection suppressed experiments under high density linear plasma dumping is ongoing.