• Energy Research
• 2021-2025close

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.

Abstract The JET 2019–2020 scientific and technological programme exploited the results of years of concerted scientific and engineering work, including the ITER-like wall (ILW: Be wall and W divertor) installed in 2010, improved diagnostic capabilities now fully available, a major neutral beam injection upgrade providing record power in 2019–2020, and tested the technical and procedural preparation for safe operation with tritium. Research along three complementary axes yielded a wealth of new results. Firstly, the JET plasma programme delivered scenarios suitable for high fusion power and alpha particle (α) physics in the coming D–T campaign (DTE2), with record sustained neutron rates, as well as plasmas for clarifying the impact of isotope mass on plasma core, edge and plasma-wall interactions, and for ITER pre-fusion power operation. The efficacy of the newly installed shattered pellet injector for mitigating disruption forces and runaway electrons was demonstrated. Secondly, research on the consequences of long-term exposure to JET-ILW plasma was completed, with emphasis on wall damage and fuel retention, and with analyses of wall materials and dust particles that will help validate assumptions and codes for design and operation of ITER and DEMO. Thirdly, the nuclear technology programme aiming to deliver maximum technological return from operations in D, T and D–T benefited from the highest D–D neutron yield in years, securing results for validating radiation transport and activation codes, and nuclear data for ITER.

Nuclear materials and energy 37, 101514 - (2023). doi:10.1016/j.nme.2023.101514 Published by Elsevier, Amsterdam [u.a.]

Nuclear materials and energy 28, 101060 - (2021). doi:10.1016/j.nme.2021.101060 Published by Elsevier, Amsterdam [u.a.]

Cold gas spraying for the production of thick tungsten (W) coatings has been investigated for use at plasma facing components in fusion devices. Since the brittle nature of W strongly impedes its deposition, a systematic study was performed using mixtures of tungsten and tantalum (Ta) powders. Whereas the use of 100% W powder was not successful yet, 2 mm thick coatings on steel were produced by using a W/Ta powder mixture with 90 vol% W, yielding a W content in the coating of 70 vol%. The coatings show negligible porosity and very good adhesion to the substrate. High heat flux experiments on samples with the size 80 × 80 mm2 were performed in order to investigate the behaviour under low (≤4 MW/m2) steady state loads and high power (∼40 MW/m2) transients. During the pulses with low power density, being typical for applications at the main chamber first wall, no defects were observed and a thermal conductivity close to that of the bulk materials was found. During the high power transients lasting for 200 ms cracks parallel to the surface appeared inside the coating.

Arcs, a source of dust particles and a localized erosion mechanism of the plasma-facing components, are found in all major fusion plasma devices. Measurements of arcs require diagnostics with high temporal and local resolution, which are not available at arc dominated locations in ASDEX Upgrade (AUG). To understand the erosion by arcing and to allow extrapolation for future fusion devices different materials are used to scan the material properties. In AUG, inserts were installed at the inner baffle region to measure the erosion by arcing. The use of polished inserts allows an accurate determination of the arc traces by depth maps obtained by laser profilometery. It turned out that the melting temperature of the materials is the main parameter for erosion. For tungsten mounted at the inner baffle, a region which is deposition dominated, an erosion rate by arcing of 1.2·1013 at cm−2 s−1 is measured. For Beryllium, 9.5·1013 at cm−2 s−1 is extrapolated from its thermal properties. As martensitic–ferritic low-activation steel is under discussion for the use in DEMO, magnetic steels were also investigated. Comparing stainless steel with magnetic steel, much deeper and wider craters are found in the latter one: they reach a depth of −80 μm. The erosion of magnetic steel by arcs is 40 times higher compared to stainless steel, which has almost the same physical properties.

Tungsten has been considered a plasma-facing material in a future fusion reactor because of its low sputtering yield and low fuel retention. It has been examined in several tokamaks. In stellarators, it has recently been used for some plasma-facing components. However, in addition to its high cost, W is difficult to machine due to its hardness and brittleness and therefore alternative materials in the form of tungsten heavy alloys are being investigated and some tests have already been performed in the ASDEX upgrade [1]. WNiFe materials are magnetic, but since magnetization saturates at ∼ 2 Tesla for W97NiFe [1], these could also be investigated for use in stellarators. Samples were prepared from pure W, W95NiCu, W97NiFe and W95NiFe alloys. The samples were exposed in the Large Helical Device (LHD) stellarator during three recent operation campaigns. The samples were inserted by means of the divertor manipulator at the positions of the strike line under H-, D- and He plasma conditions. These experiments were designed to test the samples at high thermal loads by adjusting the exposure conditions to achieve sample temperatures above, around and below the melting temperatures of Ni, Fe and Cu. During some of these exposures, although the temperatures reached above the melting limit, resulting in segregation of Ni, Fe and Cu and partial release of alloying materials, normal plasma operation continued without any radiative collapse. Scanning electron microscopy with focused ion beam (SEM/FIB), energy dispersive X-ray spectroscopy (EDX) and glow discharge optical emission spectroscopy (GDOES) measurements confirmed the observed change in surface morphology.