Vapor Shield Research Articles

Coupled model for liquid lithium plasma facing components

Numerical analysis provides the design choice and operating window of liquid metal Plasma Facing Components (PFC) concepts. Coupled analysis of boundary plasma together with the surrounding boundary structures is required. To achieve this goal, PPPL is developing a comprehensive multi-physics model for modeling of PFCs in fusion devices. The model includes the fluid-kinetic code SOLPS-ITER and the flow and heat transfer code CFX from ANSYS. SOLPS-ITER was augmented with a liquid metal boundary condition algorithm, allowing direct two-way coupling of the plasma analysis with the two-dimensional analytical slab flow model which includes heat convection in the liquid metal PFC. The target heat flux resulting from this coupled analysis is used as a boundary condition for detailed 3D Computational Fluid Dynamics (CFD) Magneto Hydro Dynamics (MHD) and heat transfer analysis. A new formulation of MHD equations is introduced in the numerical procedure ensuring current conservation of the discretized equations. Results of the 3D analysis are used for final validation of the coupled model. A PFC design where a porous wall is used to stabilize the liquid metal surface, while MHD drive is used to push the liquid metal flow inside the PFC, will be investigated in the regimes where vapor shielding is created for enhanced volumetric plasma heat dissipation.

Tuning hydrogen boil-off via inert gas insulation purge in a vapor cooled shielded fuel tank

Vapor cooled polymer composite tanks for liquid hydrogen are a newopportunity for light weight power-system integration into unmanned aerial vehicles. Vapor shielding has the ancillary benefit of pre-conditioning the hydrogen for use by a fuel cell. However, the mass flow rate required by a fuel cell should be matched by the natural boil-off rate of the tank. With limited ability to vary insulation thicknesses, the challenge becomes satisfying the fuel cell consumption in UAVs at different mass flow rates when at varying stages of flight from take-off, cruise, and landing. This paper provides experimental measurements of mass flow rate of hydrogen from the tank considering batch purging with helium or argon in the vapor cool shielded insulation layer. A 5-shell vapor cooled shielded tank is used with two layers of insulation including aerogel, helium or argon, and two vapor channels. A more complete understanding of system-level thermal conductance is determined by mass flow of hydrogen leaving the tank during hypothetical flight.

Influence of the vapor shield potential control on the post-arc sheath development in vacuum interrupters

In tank-type multi-break vacuum circuit breakers (VCBs) and gas-insulated switchgear (GIS), the non-uniform distribution of vapor shield potential in the vacuum interrupters impacts the post-arc process significantly. To explore the effect of the vapor shield potential on the development of the post-arc sheath, a particle-in-cell model with shield potential control is proposed to obtain two-dimensional distributions of post-arc particles. Based on the synthetic test circuit and probe array, the test platform of the post-arc characteristic diagnostic is set up to verify the influence of vapor shield potential distribution on the post-arc sheath. The extinguishing process of the vacuum arc is also observed by the high-speed CMOS camera. Simulation and experimental results show that at shield potential of 0.5 transient recovery voltage, the post-arc sheath develops most rapidly and the distribution of cathode spots is the most uniform. These findings can provide valuable insights for the design of vacuum interrupters and the control of shield potential in tank-type GIS.

Physics and Technology Research for Liquid-Metal Divertor Development, Focused on a Tin-Capillary Porous System Solution, at the OLMAT High Heat-Flux Facility

The operation of the Optimization of Liquid Metal Advanced Targets (OLMAT) facility began in April 2021 with the scientific objective of exposing liquid-metal plasma facing components (PFCs) to the particle and power fluxes provided by one of the hydrogen neutral beam injectors of the TJ-II stellarator. The system can deliver heat fluxes from 5 to 58 MW m−2 of high energy hydrogen neutral particles (≤ 33 keV) with fluxes up to 1022 m2 s−1 (containing an ion fraction ≤ 33% in some instances), pulsed operation of 30–150 ms duration and repetition rates up to 2 min−1. These characteristics enable OLMAT as a high heat flux (HHF) facility for PFC evaluation in terms of power exhaust capabilities, thermal fatigue and resilience to material damage. Additionally, the facility is equipped with a wide range of diagnostics that includes tools for analyzing the thermal response of the targets as well as for monitoring atomic/plasma physics phenomena. These include spectroscopy, pyrometry, electrical probing and visualization (fast and IR cameras) units. Such particularities make OLMAT a unique installation that can combine pure technological PFC research with the investigation of physical phenomena such as vapor shielding, thermal sputtering, the formation/characterization of plasma plumes with significant content of evaporated metal and the detection of impurities in front of the studied targets. Additionally, a myriad of surface characterization techniques as SEM/EDX for material characterization of the exposed PFC prototypes are available at CIEMAT. In this article, first we provide an overview of the current facility upgrade in which a high-power CW laser, that can be operated in continuous and pulsed modes (0.2–10 ms), dump and electrical (single Langmuir) probe embedded on the target surface have been installed. This laser operation will allow simulating more relevant heat loading scenarios such as nominal steady-state divertor heat fluxes (10–20 MW m−2 in continuous mode) and transients including ELM loading and disruption-like events (ms time scales and power densities up to GW m−2 range). The work later focuses on the more recent experimentation (2022 fall campaign) where a 3D printed Tungsten (W) Capillary Porous System (CPS) target, with approximated 30 μm pore size and a 37% porosity and filled with liquid tin. This porous surface was a mock-up of the PFC investigated in the ASDEX Upgrade divertor manipulator. The target composed with this element was eventually exposed to a sequence of shots with the maximum heat flux that OLMAT provides (58 ± 14 MWm−2). Key questions as resilience to dry-out and particle ejection of the liquid metal layer, its refilling, the induced damage/modification of the porous W matrix and the global performance of the component are addressed, attempting to shed light on the issues encountered with the PFC at tokamak scale testing.

Surface plasmon resonance imaging for analyzing the local variation of evaporation flux of multiple binary mixture droplets

This study investigates the local variation of concentration and selective evaporation flux affected by the droplet-to-droplet distance during the evaporation of multiple-binary mixture droplets (M-BMDs). Surface plasmon resonance imaging (SPRi) is used to measure the local liquid concentration of the binary mixture droplets and to analyze the spatial distributions of liquid concentration with the estimated local changes of surface tension during evaporation. In addition, we conduct three-dimensional numerical simulations to predict the ethanol vapor distributions and the local ethanol evaporation flux for single- and multiple-binary mixture droplets. It was found that the total evaporation time of M-BMDs is delayed due to vapor accumulation in the region adjacent to the neighboring droplets for M-BMDs, which significantly reduces the local evaporation flux. Also, the local liquid concentration of M-BMDs was found to vary, showing a lower ethanol concentration far away from the adjacent region due to vapor shielding and selective evaporation. Moreover, the scaled lifetime, which is the ratio of evaporation time between multiple droplets and a single droplet, shows good agreement with the previously reported experiments and theoretical models.

Thermomechanical Analysis of a PFC Integrating W Lattice Armour in Response to Different Plasma Scenarios Predicted in the EU-DEMO Tokamak

Despite the high performance exhibited by tungsten (W), no material would be able to withstand the huge loads expected with extreme plasma transients in EU-DEMO and future reactors, where the installation of sacrificial first wall limiters is essential to prevent excessive wall degradation. The integration of W lattices in the architecture of such components can allow for meeting their conflictual requirements: indeed, they must ensure the effective exhaust of the nominal thermal load during stationary operation; when transients occur, they must thermally insulate and decouple the surface from the heat sink, promoting prompt vapour shielding formation. Starting from the optimised layouts highlighted in a previous study, in this work, a detailed 3D finite element model was developed to analyse in depth the influence of the actual features of the latticed metamaterial on the overall performance of the EU-DEMO limiter PFC on the basis of a flat tile configuration. Its main goal is to help in identifying the most promising layout as a preconceptual design for the fabrication of a small-scale mock-up. For this purpose, the complex geometry of a W-based lattice armour was faithfully reproduced in the model and analysed. This allowed for a detailed assessment of the thermally induced stresses that develop in the component because of the temperature field in response to a number of plasma scenarios—above all, normal operation and ramp down. Structural integrity was verified through the acceptance criteria established for ITER. The two optimised layouts proposed for the PFC were able to effectively meet the requirements under normal reactor operating conditions, while they missed some requirements in the ramp-down case. However, the first HHF tests will be performed in order to benchmark the analyses.

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.

Effect of continuously flowing liquid Li limiter on particle and heat fluxes during H-mode discharges in EAST

• Particle and heat fluxes were successfully controlled by using continuously flowing liquid Li (FLiLi) limiter in the H mode discharges with high plasma heating power in EAST. • By using flowing liquid Li limiter, the fuel recycling could effectively reduce by ∼ 50% and Zeff could decrease from 2.3 to 1.6. • The decreased recycling and impurity radiation led to the achievement of high energy confinement plasma with the average stored energy increased up to ∼ 290 kJ. • Due to Li vapor shielding effect, ∼40% plasma heat flux was dissipated before arrived at the Li limiter. Particle and heat fluxes were successfully controlled by using a continuously flowing liquid Li (FLiLi) limiter in the H-mode discharges with high plasma heating power in the Experimental Advanced Superconducting Tokamak device. There were strong interactions between the FLiLi limiter and high-power plasma with a ∼ 8.3 MW source heating power, and successively, a bright Li radiation ring was produced, which effectively decreased fuel particle recycling by approximately 50%. Due to Li efflux from FLiLi during a series of high-power discharges, an obvious real-time wall conditioning effect was produced, and fuel particle recycling further decreased. Moreover, the value of Z eff decreased from 2.3 to 1.6 due to a decrease in impurity sources; this was attributed to the accumulation of Li deposited on the first wall, which effectively protected the wall materials. The decreased recycling and impurity radiation achieved high-energy confinement plasma, and the average stored energy increased up to ∼ 290 kJ. Moreover, due to the effect of Li vapor shielding, nearly 30% plasma heat flux was dissipated before it arrived at the Li limiter. These results promote further exploration of liquid Li solutions for the critical challenge of heat flux handling and particle control in fusion power plants.

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.

Plasma facing components with capillary porous system and liquid metal coolant flow

Liquid metal can create a renewable protective surface on plasma facing components (PFC), with an additional advantage of deuterium pumping and the prospect of tritium extraction if liquid lithium (LL) is used and maintained below 450 °C, the temperature above which LL vapor pressure begins to contaminate the plasma. LM can also be utilized as an efficient coolant, driven by the Lorentz force created with the help of the magnetic field in fusion devices. Capillary porous systems can serve as a conduit of LM and simultaneously provide stabilization of the LM flow, protecting against spills into the plasma. Recently, a combination of a fast-flowing LM cooling system with a porous plasma facing wall (CPSF) was investigated [A. Khodak and R. Maingi, Nucl. Mater. Energy 26, 100935 (2021)]. The system takes an advantage of a magnetohydrodynamics velocity profile as well as attractive LM properties to promote efficient heat transfer from the plasma to the LL at low pumping energy cost, relative to the incident heat flux on the PFC. In the case of a disruption leading to excessive heat flux from the plasma to the LM PFCs, LL evaporation can stabilize the PFC surface temperature, due to high evaporation heat and apparent vapor shielding. The proposed CPSF was optimized analytically for the conditions of a fusion nuclear science facility [Kessel et al., Fusion Sci. Technol. 75, 886 (2019)]: 10 T toroidal field and 10 MW/m2 peak incident heat flux. Computational fluid dynamics analysis confirmed that a CPSF system with 2.5 mm square channels can pump enough LL so that no additional coolant is needed.

Analytical expressions for thermophysical properties of solid and liquid beryllium relevant for fusion applications

The status of the literature is reviewed for thermophysical properties of pure polycrystalline solid and liquid beryllium which constitute input for the modeling of intense plasma–surface interaction phenomena that are important for fusion applications (thermal analysis, vapor shielding, melt motion, arcing, dust generation, dust transport). Reliable experimental data are analyzed for the latent heats, specific isobaric heat capacity, electrical resistivity, thermal conductivity, mass density, vapor pressure, work function, total hemispherical emissivity and absolute thermoelectric power from the room temperature up to the normal boiling point of beryllium as well as for the surface tension and the dynamic viscosity across the liquid state. Analytical expressions are recommended for the temperature dependence of these thermophysical properties, which involve high temperature extrapolations given the absence of extended liquid beryllium measurements.

Recent progress, liquid metal use as plasma facing component and vapor shielding of high heat flux

Recent progress, liquid metal use as plasma facing component and vapor shielding of high heat flux

SOLPS-ITER simulations of a CPS-based liquid metal divertor for the EU DEMO: Li vs Sn

In this work, we study the effect of installing a liquid metal divertor (LMD) using a capillary-porous structure in the EU DEMO tokamak within the same envelope of the baseline solid divertor. We used the SOLPS-ITER code to model the scrape-off layer (SOL) plasma and neutrals, coupled to a target thermal model to enable the self-consistent calculation of the LM target erosion rate, and adopting a fluid neutral model for the sake of simplicity. First calculations considering only D and Li (or Sn) showed a significant reduction of the steady state target heat load with respect to simulations considering only D, thanks to vapor shielding. Nevertheless, the computed peak target heat flux (∼31 MW m−2 and ∼44 MW m−2 for Li and Sn, respectively) was still larger than/borderline to the power handling limit of the LMD concepts considered. Moreover, the impurity concentration in the pedestal—a proxy for the core plasma dilution/contamination—was computed to be above/close to tolerability limits suggested by previous COREDIV calculations. These results indicate that the operational window of an LMD for the EU DEMO, without any additional impurity seeding, might be too narrow, if it exists, and that Sn looks more promising than Li. A second set of calculations was then performed simulating Ar seeding in the SOL, to further reduce the target heat load, and consequently the metal erosion rate. It was found that the mitigation of the plasma heat load due to Ar radiation in the SOL effectively replaces the radiation associated to vapor shielding in front of the target, thus allowing to operate the LMD in a regime of low target erosion. The resulting operational window was found to be significantly wider, both in terms of tolerable peak target heat flux and of acceptable core plasma contamination.

Conceptual design of reciprocating probes and material-testing manipulator for tokamak COMPASS Upgrade

Three new in-vessel manipulators are designed and built for the new COMPASS Upgrade tokamak with uniquely high vessel temperature (250–500 °C) and heat flux density (perpendicular to divertor surface q ⊥ ∼ 80 MW/m2 and q ∥ ∼ GW/m2 at separatrix), which challenges the edge plasma diagnostics. Here we show their detailed engineering designs supported by heat conduction and mechanical models. Deep reciprocation of electrostatic probes near the separatrix should be possible by optimizing older concepts in (a) the head and probe geometry, (b) strongly increasing the deceleration up to 100× gravity by springs and strengthening the manipulator mechanical structure. One reciprocates close to the region of edge plasma influx (the outer midplane), the other at the plasma sink (between the outer divertor strike point and X-point), for studying the plasma divertor (impurity-seeded) detachment and liquid metal vapor transport. Both probe heads are equipped with a set of ball-pen and Langmuir probes, measuring reliably and extremely fast (10−6 s) local (1 mm resolution) plasma potential, density, electron temperature and heat flux and even ion temperature with 10−5 s resolution. The divertor manipulator (without reciprocation) will place various material test targets at the outer divertor. Unique will be its capability to increase 15× the surface heat flux with respect to the surrounding tungsten tiles just by controllable surface inclination of the test targets. We plan to test liquid metal targets where such inclined surface was found critical to achieve the desired mode with lithium vapor shielding. Even in the conservative expected performance of COMPASS Upgrade, we predict to reach and survive the EU DEMO relevant heat fluxes.

Effect of lithium vapour shielding on hydrogen plasma parameters

A liquid Li vapour-box divertor is an attractive heat exhaust solution for future fusion reactors. Previous works have established the ability of vapour shielding to protect the wall, but it has not been possible to directly determine the effects of Li vapour on the plasma parameters. Experiments to investigate this were carried out in Magnum-PSI, which is able to generate a plasma with DEMO-divertor relevant conditions. 3D printed tungsten capillary porous structures filled with Li have been used as targets. A reciprocating Langmuir probe was used to determine electron temperature and density close to the target, while the power reduction to the coolant due to vapour shielding was increased from 0% to 50%. The Langmuir probe measurements directly determined an increase of density by up to 50% while electron temperature could be inferred to have dropped by up to 33% compared to the solid target reference case.

Influence of lithium and silicon coatings on tungsten erosion in arc plasma

• Arc erosion of W substrate can be effectively mitigated by Li and Si coatings. • Selective formation of spots on coatings is responsible for erosion mitigation. • Vapor shielding of Li coating provides effective mitigation of tungsten erosion. Arc spots formed on a plasma-facing component (PFC) have a stunning resemblance with the arc spots formed on a cathode of a vacuum arc. Hence, in this paper, a laboratory-scale vacuum arc process is used to get a better understanding of the arc erosion process of the tungsten (W) material. The performance of lithium (Li) and silicon (Si) coatings in erosion mitigation of W material is analyzed. The virgin W substrate was eroded significantly by arc spots and numerous W droplets were ejected into the interelectrode plasma from cathode with velocity of tens m/s. However, the W substrates can be effectively protected by Li and Si coatings from arc erosion. Preferential formation of arc spots on the coatings is responsible for the arc erosion mitigation. Compared with the virgin W substrate, no ejection of W droplets from the cathode was observed in the cases of coatings. Li and Si droplets were accidently ejected from the cathode and the ejected droplets were rapidly ablated in the arc plasma. For the case of Si coating, the W substrate was slightly melted and eroded due to the arc spots formed on the coating. By contrast, no obvious melting and erosion were observed on the W substrate for the case of Li coating, which is possibly due to the vapor shielding effect of Li between the W substrate and arc spots. These results confirm coatings served as a sacrificial protective layer and can reduce tungsten surface erosion caused by arcs, which provides a new reference for protecting first wall materials in future fusion devices.

Vapour shielding of liquid-metal CPS-based targets under ELM-like and disruption transient loading

This paper presents experimental studies of plasma-surface interactions during powerful plasma impacts of a quasi-stationary plasma accelerator (QSPA) on the Sn capillary porous systems (CPSs) in conditions simulating disruption and edge localized modes (ELM) like loads. Experiments were carried out using two QSPA devices. ELM-like plasma exposures were performed with QSPA-M test-bed facility. A large-scale QSPA Kh-50 device was used to simulate plasma disruptions and giant ELMs. Variation of the plasma stream energy density has been performed to study the onset of vapour shield. It is shown that during plasma exposures of a Sn-CPS target with the QSPA plasma load <1 MJ m−2, single dust particles traces have been registered. A further increase in the heat load leads to the splashing of the eroded material. For ELM-like impacts, a rather weak melt motion was observed on the target surface. A post-mortem analysis has shown that the CPS structure was not destroyed in the course of many repetitive ELM-like pulses. Surface morphology has changed from a smooth surface to corrugation structures with the formation of some cavities in mesh cells due to the influence of the surface tension and capillary effects. Spectral lines of Sn I and Sn II have been identified by optical emission spectroscopy in the near-surface plasma. A plasma shield, that consists mostly of Sn neutrals appears at Q ∼ 0.1 MJ m−2. An increase in the surface heat load resulted in the intensive emission of Sn II lines, which started to be observed at Q ∼ 0.3 MJ m−2. The plasma electron density near the surface increases significantly at Q > 0.5 MJ m−2, which corresponds to the strong vapour shielding of the exposed surface. A comparison between the obtained results on the vapour shielding of Sn CPS and available numerical simulation using the TOKES code has been performed.

Pulse Shape Dependence of Vapor Shielding Efficiencies During Transient Heat Loads

Pulse Shape Dependence of Vapor Shielding Efficiencies During Transient Heat Loads

Conceptual design of a liquid-metal divertor for the European DEMO

Liquid metal (LM) divertors are considered for the European DEMO reactor, because they may offer improved performance compared to the tungsten monoblock concept. The goal of this work is to provide a concept design, and explore the limitations of liquid metal divertors. To this end, a set of design requirements was formulated in close collaboration with the EUROfusion Power Plant Physics and Technology team (responsible for the design of the EU-DEMO). Tin was chosen as the preferred liquid metal, because unacceptable Tritium retention issues arise when lithium is used in DEMO. A concept design was then chosen that consists of water cooled pipes that are square on the outside and round on the inside, a corrosion barrier, and a 3D-printed porous tungsten armor layer filled with liquid tin. The porous armor layer acts as a Capillary Porous System (CPS). The design was analyzed using thermo-mechanical FEM simulations for various armor thicknesses and heat sink materials: Densimet, W/Cu composites, and CuCrZr. The highest heat loading capability achieved is 26.5 MW/m2 in steady state (18.9 MW/m2 when taking into account a safety margin of 1.4). This is achieved using a CuCrZr pipe, with a 1.9 mm thick armor. When increasing the armor layer to 3 mm thick, more than 80 MW/m2 can be withstood during slow transients thanks to vapor shielding, but at the same time the steady-state capability is reduced to 18 MW/m2. Resilience against disruptions cannot yet be proven, but is deemed within the realm of possibility based on estimates regarding the behavior of vapor shielding. This should be further investigated. Overall, the concept is considered a significant improvement compared to the original specifications (which are also the specifications to the tungsten monoblocks: 10 MW/m2 in steady state, and ∼20 MW/m2 during slow transients). Moreover, the possibility of withstanding disruptions is regarded as a potentially major improvement.

Performance of liquid-lithium-filled 3D-printed tungsten divertor targets under deuterium loading with ELM-like pulses in Magnum-PSI

A fusion reactor divertor must withstand heat flux densities <10 MW m−2. Additionally, it may have to withstand millisecond pulses on the order of 0.5 to 30 MJ m−2 due to (mitigated) edge-localized modes (ELM) occurring with 30 to 60 Hz. We investigate if these requirements can be met by capillary porous system (CPS) liquid lithium divertors (LLD). 3D-printed tungsten CPS targets were exposed in the linear plasma device Magnum-PSI, to deuterium plasma discharges lasting 15 s, generating 1.5 to 16 MW m−2, and T e ∼ 1.5 eV. Additionally, ELM-like pulses were superimposed on top of the steady state for 3 s with a frequency of 2 and 100 Hz, power flux densities of 0.3 to 1 GW m−2, and T e up to ∼14 eV. All Li targets survived without damage. The surface temperature (T s) was locked at ∼850 °C, which was attributed to power dissipation via vapor shielding. Meanwhile, unfilled reference targets melted during the severest pulsed loading. A blue grayish layer of presumably LiD formed when T s < 500 °C, but disappeared when the locking temperature was reached. This implies that LiD formation can be avoided, but that it may require a surface temperature at which Li evaporation excessively contaminates the core plasma in a tokamak. During pulsed loading the plasma facing surface remained wetted in all conditions. Bolometry indicated that, only during pulses, there was a large increase in radiative power dissipation compared to targets without Li. A high speed camera with a Li-I filter showed that strong Li evaporation continued up to 5 ms after a pulse. Overall, the liquid-lithium-filled 3D-printed tungsten targets were found to be highly robust, and 3D-printing can be considered as a promising manufacturing technique for LLDs. Further research is needed particularly on the formation of LiD and the associated tritium retention, as well as the impact of enhanced evaporation during and after ELMs on the core plasma.