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Liquid metal divertors are discussed as a potential alternative to divertor components made entirely of solid materials such as tungsten. However, liquid metal interaction with H isotope plasmas poses entirely new challenges, including the stability of the liquid surface. In this article, the interaction of low-temperature D plasmas with both free liquid tin surfaces, and the liquid metal confined in a capillary porous system (CPS) is presented. Free surfaces showed intolerable instabilities due to formation of gas bubbles in the melt, as opposed to a strong stabilizing effect of the CPS. A tendency of liquid tin to well up and entirely cover the surface of the CPS may be beneficial in terms of quickly re-establishing a tin surface after depletion by transient overloading. On the other hand, this appears to counteract efforts to fully suppress superficial bubble formation and bursting, and the resulting ejection of tin droplets by designing a CPS with very small pores.

Using liquid metals confined in capillary porous structures (CPSs) as a plasma-facing component (PFC) could prolong the lifetime of the divertor in the high heat flux area. However, the high atomic number of tin (Sn) limits its acceptable fraction in the main plasma. Therefore, a crucial step in developing this concept is to test it in a tokamak environment, particularly in the diverted plasma region, e.g. ASDEX Upgrade (AUG). In this paper, the design of liquid tin module (LTM) is explained, and the testing in the high heat flux device GLADIS before its use in AUG is presented. The LTM was additively manufactured using selective laser melting, consisting of a 1.5mm porous layer tungsten (W) directly attached to a solid W bulk. The LTM has a plasma-facing area of 16×40mm2 and was filled with 1.54g of Sn. In GLADIS, the module was exposed to power loads between 2 and 8MWm−2 for 1 up to 10s, first unfilled and later filled with Sn. The surface temperature was monitored with infrared imaging and pyrometry. The thermal response was used to compare with simulations in Ansys Mechanical, enabling a determination of the module’s effective thermal properties. Sn droplets could be observed on the infrared camera, until a surface temperature of about a 1000°C was reached. The enhanced wetting of tin on the plasma-facing surface, which was observed by a visible camera, suggests that there is a conditioning of the surface, possibly due to the removal of impurities and oxides. Subsequent examinations of the adjacent tile revealed minor Sn leakages emanating from the module’s edge. Furthermore, the module showed no indication of mechanical failure. Therefore, these results indicated that the LTM qualifies for the heat fluxes expected in ASDEX Upgrade.

One of the ways to extend the lifetime of the divertor for DEMO could be to replace the solid tungsten plasma-facing components with liquid tin (Sn) confined in a tungsten capillary porous structure (CPS). Testing a CPS in a divertor plasma configuration is crucial for the development of a liquid metal divertor (LMD) to understand how the main plasma is affected. Only a limited Sn concentration is allowed in the plasma core, due to the high radiative losses associated with the high atomic number of Sn (50). Therefore, it is necessary to test a small-scale LMD filled with Sn in a tokamak environment, which has not previously been done. In ASDEX Upgrade, a liquid tin module (LTM) has been exposed by means of the divertor manipulator. During plasma flat-top, the outer strike point (OSP) was placed onto the pre-heated LTM and held there for a time interval between 2 and 3.4s over multiple discharges. Photographs of the LTM taken after each discharge, revealed macroscopic Sn leakage onto the adjacent tile. Simulations with the HeatLMD code predicted an acceptable tin erosion near the LTM with thermal sputtering dominating over evaporation. However, spectroscopic measurements revealed an order of magnitude higher erosion. Since this remained constant when the OSP was held on the LTM so that the surface temperature increased, evaporation could be excluded as the main source of Sn erosion. Comparison between discharges with different durations of OSP location on the LTM revealed an increase in core radiation up to 1.5MW due to Sn. The 1.5D-impurity transport code STRAHL was used to interpret this increase in total plasma radiation and revealed a Sn concentration in the main plasma of up to 1.4×10-4. Given that the LTM only covered about 1/650 of the outer divertor circumference, extrapolating to a full toroidal divertor implies erosion is above acceptable limits. The unexpectedly high Sn fraction in the main plasma is attributed to the ejection of Sn droplets reaching the main plasma, which may have originated from either the CPS or leaked tin. This conclusion is also supported by splashes of tin droplets, which were observed on the adjacent divertor tile and one ∼0.5m downstream. Therefore, to make a Sn-filled LMD a viable alternative to solid tungsten, the formation of droplets must be reduced by two orders of magnitude.

We developed a novel, recently patented application of hydrogenography, i.e., the change of the optical properties of certain indicator materials (e.g., yttrium) when these are charged with hydrogen isotopes. By applying patterned, discontinuous yttrium films to the back side of permeation samples, we circumvented the issue that the hydrogen isotopes could also diffuse laterally within the yttrium film, which might blur or even prevent detection of laterally inhomogeneous diffusion effects through the actual sample. Thus, we were able to directly visualize the preferential permeation of deuterium along grain boundaries in recrystallized tungsten foils exposed to a low-temperature deuterium plasma. Similarly, we were able to show that permeation of deuterium focuses on the percolating Fe–Ni matrix phase of the tungsten heavy alloy HPM 1850. These findings give a strong incentive to assess whether 1-D diffusion simulations, which implicitly assume lateral homogeneity, are generally sufficient to accurately describe H isotope diffusion in fusion reactor materials. At least under certain conditions, an approach considering the effects of internal 3-D structures of the materials – like those described here – may be necessary.

Recrystallization is one of the most pronounced microstructural changes of tungsten when subjected to cyclic high heat flux (HHF) loads. In the framework of the European DEMO divertor development an intensive R&D programme is being performed. Many HHF tests of tungsten monoblocks have shown that recrystallization and grain growth have been deemed a major degradation feature leading to brittleness and reduced strength. In the previous recrystallization studies, tungsten was normally heat-treated in a furnace in slow uniform heating. However, in a HHF test or in the assumed fusion reactor operation, the water-cooled tungsten armor is rapidly heated by cyclic HHF pulses generating a steep temperature gradient (20 MW/m2 loading results in about 200 K/mm) and thermal stresses. This difference raises the question as to whether a furnace heat-treatment condition properly simulates the cyclic HHF loading case in terms of recrystallization behaviour if the heat exposure conditions are kept comparable to each other. The present paper addresses this issue. To this end, a comparative microstructural study was performed for two different groups of tungsten samples: one tested under well-defined furnace heat-treatment conditions (1500 ℃, 2100 ℃) and the other one taken from the monoblocks (at the positions of the same corresponding temperatures) of a water-cooled mock-up tested under HHF loads at 20 MW/m2, 500 cycles. Extensive quantitative image analysis was carried out based on detailed microstructural and crystallographic characterization and micro-hardness was measured. The HHF loaded surface is dominated by the formation of extremely large grains in contrast to furnace heating at the same temperature. The samples heated at 2100 °C exhibited a remarkable difference in recrystallization and grain growth behaviour and hardness values between the two heating cases. Two commercial tungsten grades (AT&M, ALMT) showed a similar behaviour to each other.

Advanced materials such as tungsten fibre-reinforced composites allow to overcome severe weaknesses of the baseline materials for plasma-facing components — copper and tungsten. The effect of the fusion environment on the mechanical properties of these materials, e.g. the embrittlement by neutron irradiation, plays a key role for the development of future fusion reactors. To simulate this effect, high-energy ions are used as a substitute for the displacement damage by neutrons. We propose the use of very fine tungsten wire as a possibility of studying the influence of irradiation damage on the mechanical properties. This is possible as they allow full-depth irradiation of almost the entire volume despite the limited penetration depth of ions. Geometrical size effects are mitigated due to the nanoscale microstructure of the wire. In addition, similar wire is used in tungsten fibre-reinforced composites. Thus, the investigation of irradiated wire can directly be used for the prediction of the bulk composite properties. For the proof of this concept tungsten wire with a diameter of 16 μm was electrochemically thinned to 5 μm and irradiated with 20.5 MeV W6+ions. The mechanical properties were subsequently determined by macroscopic tensile testing. Irradiation to 0.3, 1 and 9 dpa did not lead to a change of the mechanical behaviour. Both strength and ductility, the latter indicated by the reduction of area, were similar to the as-fabricated state.

The exposure of metal surfaces to energetic particles originating for example from a plasma can lead to the formation of blisters. These features are caused by hydrogen-filled gas bubbles near the surface and can influence the diffusion and retention of hydrogen atoms in the bulk material during plasma exposure. No comprehensive theory of the underlying processes governing these effects has been developed so far. A major constraint to the investigation of blisters has been the effort that is required to identify and characterize a statistically relevant ensemble of blisters. Therefore, the analysis of a limited number of blisters in isolated observations before and after plasma exposure has been state of the art so far.In this article, a framework is presented that allows to automatically analyze blisters in image data with the aid of convolutional neural networks (CNNs) that are trained with artificial training data. It thereby resolves the limitations of existing analysis methods that prevented the large-scale and in-situ investigation of blistering up to now. Automated routines are used to identify and localize blisters, estimate important blister parameters such as size, and track individual blisters over time. Apart from discussing the principles underlying this approach and describing the implementation of the framework in detail, this article also presents first results acquired with a molybdenum sample that was exposed to a low-temperature deuterium plasma. The results indicate a blister identification accuracy of the framework of more than 80% within the considered parameter ranges and overall demonstrate the feasibility of the approach in general.

In future fusion reactors, plasma-facing materials (PFMs) have to withstand unique conditions such as high temperatures, ion and neutron irradiation. Tungsten (W) has been established as main candidate material due to its favorable properties regarding the fusion environment but brings one major challenge: Its brittleness at moderate temperatures can lead to failure of tungsten components. Tungsten fiber-reinforced tungsten (Wf/W), a tungsten matrix containing drawn tungsten fibers, was developed to mitigate this problem by using extrinsic toughening mechanisms to achieve pseudo-ductility. The deuterium (D) retention in Wf/W manufactured by chemical vapor deposition (CVD) has been investigated using Wf/W single layered model systems consisting of a single plane of unidirectional tungsten fibers embedded in a tungsten matrix produced by CVD. Various parameters with potential influence on the D retention, such as the choice of an erbium oxide interface and potassium doping, have been included in the investigation. The samples have been ground to varying distances between surface and fiber plane - exposing distinct details of the Wf/W microstructures at the surface. The samples were exposed to a low temperature D plasma at 370 K for 72 h resulting in a total fluence of 1025 D/m2. The D retention of all samples was measured by nuclear reaction analysis (NRA) and thermal desorption spectroscopy (TDS). The D retention in Wf/W composites is higher than in reference samples made from hot-rolled W by factors between 2 and 5. In addition, a comparison of NRA and TDS data indicates that D penetrates faster into the depth of Wf/W material than into hot-rolled tungsten.

The tungsten “heavy alloy” HPM 1850, a liquid-phase sintered composite material with two weight percent Ni and one weight percent Fe, as well as the self-passivating tungsten alloy W-10Cr-0.5Y, a high temperature oxidation resistant alloy with 10 weight percent of Cr and 0.5 weight percent of Y, were investigated with respect to their deuterium retention. The samples were deuterium loaded in an electron cyclotron resonance plasma up to a fluence of 1025m−2. The deuterium retention was then investigated by Nuclear Reaction Analysis and by Thermal Desorption. In HPM 1850 the observed deuterium amount was similar to pure tungsten, however the outgassing behaviour during thermal desorption was considerably faster. In W-10Cr-0.5Y the released deuterium amount during thermal desorption was about one order of magnitude higher; by comparison of nuclear reaction analysis and thermal desorption this was attributed to deeper diffusion of deuterium into the bulk of the material. Keywords: Deuterium retention, Tungsten heavy alloy, Tungsten self-passivating alloy, Nuclear reaction analysis, Thermal desorption

By irradiating a tungsten sample for transmission electron microscopy with deuterium plasma at 230 K, we created a high density of blisters below the plasma-exposed surface. Blisters were even found in the electron-transparent region close to the central perforation. We determined the abundance of blisters depending on the local sample thickness and found that no blisters were formed for thicknesses below about 45 nm for these specific plasma exposure conditions. After removing the approximately 10 nm thick, defect-rich layer at the plasma-exposed surface, which corresponds roughly to the ion implantation zone, by electrochemical polishing, we were able to clearly identify material distortions and dislocation network halos corresponding to blisters by scanning transmission electron microscopy. Compared with unirradiated tungsten of the same grade, we found that in the blister zone the average dislocation density is about 2 orders of magnitude higher, which may, in addition to gas enclosed in blister cavities, explain enhanced deuterium retention in blistered tungsten samples. The proof-of-principle experiments described in this article have the potential to provide constraints for theoretical models for blister nucleation. They also pave the way for direct investigations of the spatial correlation of deuterium-plasma-induced blisters and intrinsic defects. Keywords: Tungsten, Deuterium, Plasma exposure, Scanning transmission electron microscopy, Blister nucleation conditions, Defect creation

Pure and Y2O3-doped tungsten samples were simultaneously exposed to deuterium (D) plasma. The following exposure parameters were investigated: ion energy of 38 eV/D, sample temperatures of 370, 450 and 570 K, and incident fluences of 6 × 1024 and 7.5 × 1025 D/m2. The deuterium retention behavior of these tungsten materials was investigated by nuclear reaction analysis and thermal desorption spectroscopy. After exposure at 450 K Y2O3-doped tungsten shows a D depth profile comparable with that of pure tungsten. But at each investigated exposure condition Y2O3-doped tungsten shows a higher total D inventory than pure tungsten. This is attributed to the presence of intrinsic defects with high trapping energies. Investigation of the D retention behavior of four different Y2O3-doped tungsten materials, which contain the same amount of Y2O3, but have various microstructures, suggests that sub-micron pores are the underlying intrinsic defects responsible for the appearance of a high temperature D release peak. Keywords: Tungsten, Tungsten alloys, Deuterium retention, Plasma-surface interaction, Morphology, Thermal desorption spectroscopy

A method to measure deuterium permeation through tungsten near room temperature under plasma loading is presented. The permeating deuterium is accumulated in a getter layer of zirconium, titanium or erbium, respectively, on the unexposed side of the sample. Subsequently, the amount of deuterium in the getter is measured ex-situ using nuclear reaction analysis. A cover layer system on the getter prevents direct loading of the getter with deuterium from the gas phase during plasma loading. In addition, it enables the distinction of deuterium in the getter and at the cover surface. The method appears promising to add additional permeation measurement capabilities to deuterium retention experiments, also in other plasma devices, without the need for a complex in-situ permeation measurement setup. Keywords: Deuterium, Plasma, Permeation, Tungsten, Getter, Ion-beam

Up to now, analyzing the production of dislocation-type defects in the subsurface region of plasma or ion-exposed tungsten samples has been hampered by the challenging production of suitable cross-section samples for transmission electron microscopy. We present two reliable methods based on precision electropolishing to prepare cross-sections of tungsten that allow direct imaging of dislocation-type defects by scanning as well as by transmission electron microscopy. Using these methods, we are able to demonstrate a clear enhancement of the dislocation density in the caps of blisters on tungsten exposed to H isotope plasma, i.e., of surface morphologies that are correlated to subsurface cavities. As a benchmark, we also show a cross-section of tungsten irradiated by 20 MeV W6+ ions. Keywords: Tungsten, Plasma, Blistering, Cross-section, Electropolishing, Dislocations

DEMO is the name for the first stage prototype fusion reactor considered to be the next step after ITER. For the realization of fusion energy especially materials questions pose a significant challenge already today. Advanced materials solution are under discussion in order to allow operation under reactor conditions [1] and are already under development used in the next step devices. Apart from issues related to material properties such as strength, ductility, resistance against melting and cracking one of the major issues to be tackled is the interaction with the fusion plasma. Advanced tungsten (W) materials as discussed below do not necessarily add additional lifetime issues, they will, however, add concerns related to erosion or surface morphology changes due to preferential sputtering. Retention of fuel and exhaust species are one of the main concerns. Retention of hydrogen will be one of the major issues to be solved in advanced materials as especially composites and alloys will introduce new hydrogen interactions mechanisms. Initial calculations show these mechanisms. Especially for Helium as the main impurity species material issues arise related to surfaces modification and embrittlement. Solutions are proposed to mitigate effects on material properties and introduce new release mechanisms.