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Currently, tungsten fibre-reinforced (Wf) composites are regarded as promising materials for plasma-facing components of future magnetic confinement fusion devices. In this context, tungsten fibre-reinforced tungsten (Wf/W) is being investigated as a pseudo-ductile composite material overcoming the intrinsic brittleness of bulk tungsten while tungsten fibre-reinforced copper (Wf/Cu) is being developed as a high-strength composite heat sink material. In this contribution, we discuss the current development status and the progress that has been achieved recently with respect to characterization and upscaling of the aforementioned materials.In cooperation with industry, upscaling of multifilamentary W yarn fabrication was demonstrated. Multilayered W fibre braids were made from such yarns and used for the manufacturing of 400mm long medium-scale tungsten fibre-reinforced copper heat sink tubes. The maturity of short tungsten fibre-reinforced tungsten composites produced by powder metallurgy allowed the fabrication of flat tile mock-ups. Test procedure and first results of high heat flux tests are shown. Finally, we discuss the challenges and the benefits of these composites for the use in high heat flux components.

Wendelstein 7-X, the world’s largest superconducting stellarator in Greifswald (Germany), started plasma experiments with a water-cooled plasma-facing wall in 2022, allowing for long pulse operation. In parallel, a project was launched in 2021 to develop a W based divertor, replacing the current CFC divertor, to demonstrate plasma performance of a stellarator with a reactor relevant plasma facing materials with low tritium retention. The project consists of two tasks: Based on experience from the previous experimental campaigns and improved physics modelling, the geometry of the plasma-facing surface of the divertor and baffles is optimized to prevent overloads and to improve exhaust. In parallel, the manufacturing technology for a W based target module is qualified.This paper gives a status update of project. It focusses on the conceptual design of a W based target module, the manufacturing technology and its qualification, which is conducted in the framework of the EUROfusion funded WPDIV program. A flat tile design in which a target module is made of a single target element is pursued. The technology must allow for moderate curvatures of the plasma-facing surface to follow the magnetic field lines. The target element is designed for steady state heat loads of 10 MW/m2 (as for the CFC divertor). Target modules of a similar size and weight as for the CFC divertor are assumed (approx.

The standard positive triangularity H-mode tokamak operation with Single Null Divertor (SND) configuration and tungsten monoblocks targets can face some difficulties in providing a solution scalable towards the realization of the fusion reactor. To provide a safer solution, the adoption of alternative divertor magnetic configurations (ADCs) has been considered to provide on the targets a stationary heat load of less than 10 MW/m2 presently considered an upper technological limit for a nuclear fusion reactor and to reduce enough plasma temperature to avoid tungsten sputtering and its possible plasma contamination. Additionally, different plasma scenarios have been considered to avoid the huge transient energy released due to the type-I ELMs of the high confinement H-mode operation.The new high field superconducting divertor tokamak test facility (DTT) [1] is presently under construction to specifically study power exhaust solutions in reactor relevant regimes. The first DTT divertor will use the ITER-like technology based on full tungsten monoblocks bonded on CuCrZr cooling tubes. To test and compare a wide set of different power exhaust solutions, the divertor is being designed to be compatible with the standard single null (SND) configuration but also with some ADCs, like the X divertor (XD) and the hybrid Super-X/long leg SN. Additionally, the l-mode negative triangularity (NT) operation is considered important to explore as a solution to avoid ELMs.Two different closed divertors with the same grazing angle (α = 2°) for the reference SND configuration have been considered: a so called “narrow” one with normalized divertor parameters similar to those presently foreseen for DEMO divertor and the “wide” one with a 50 % wider target aperture which allows testing long legs configurations as well as snowflake divertor configurations (SFD). For the “wide” divertor different dome sizes have been also considered.The comparison between different shapes and their optimization have been done with the 2D edge fluid-kinetic code SOLEDGE2D-EIRENE. The code easily manages all magnetic configurations and can resolve the heat load on all components thanks the mesh filling the whole divertor and vessel volume.In a representative sub-set of all the possible magnetic equilibrium configurations that can be realized in DTT, the edge modelling shows a better performance of the “wide” divertor compared to the “narrow” one. In fact, the “wide” divertor allows to have lower plasma temperature at the targets in pure deuterium at low PSOL power and also lower impurity concentrations is required to achieve plasma detachment by impurity seeding at the reference full power DTT scenario. The results seem to indicate that the improvement due to the longer legs possible with the “wide” divertor can overcome its lower apparent closure compared to the “narrow” divertor solution.

The divertor target plates are the most thermally loaded in-vessel components in a fusion reactor where high heat fluxes are produced on the plasma-facing components (PFCs) by intense plasma bombardment, radiation and nuclear heating. For reliable exhaust of huge thermal power, robust and durable divertor target PFCs with a sufficiently large heat removal capability and lifetime has to be developed. Since 2014 in the framework of the preconceptual design activities of the EUROfusion DEMO project, integrated R&D efforts have been made in the subproject ‘Target development’ of the work package ‘Divertor’ to develop divertor target PFCs for DEMO. Recently, the first R&D phase was concluded where six (partly novel) target PFC concepts were developed and evaluated by means of non-destructive inspections and high-heat-flux fatigue testing. In this paper, the major achievements of the first phase activities in this subproject are presented focusing on the design rationales of the target PFC concepts, technology options employed for small-scale mock-up fabrication and the results of the first round high-heat-flux qualification test campaign. It is reported that the mock-ups of three PFC concepts survived up to 500 loading cycles at 20 MW/m² (with hot water cooling at 130 °C) without any discernable indication of degradation in performance or structural integrity. Keywords: DEMO, Divertor targets, High-heat-flux, Plasma-facing component, Composites, EUROfusion

The feasibility of getting “D” shaped plasma configurations in FTU, with a possible X point close to the first wall, has been investigated with the aim of achieving an H-mode regime in this machine. This regime could allow both evaluating the thermal effects on the liquid lithium limiter due to the possible Edge Localized Modes and studying the L-H transition properties in low recycling conditions due to the presence of lithium.. An alternative design for the magnetic system in FTU has been also proposed, to realize an X-point inside the plasma chamber, close to the Liquid Lithium Limiter.Preliminary experiments with elongated configurations and limited ECRH additional heating power did not allowed approaching the L-H transition but they were used to develop a proper elongation control. This controller allowed guaranteeing the vertical stability in elongated configurations despite the reduced power available for the horizontal field coils in FTU. The elongation was stably keep over 1.2, while the lithium limiter was very close to the last close flux surface. Neither limiter damages nor plasma pollution were observed in these configurations.A possible alternative connection of the poloidal field coils in FTU is here proposed, with the aim of achieving a true X-point configuration with a magnetic single null well inside the plasma chamber and strike points on the lithium limiter. A preliminary assessment of this design allowed estimating the required power supply upgrade and showed its compatibility with the existing mechanical structure and cooling system, at least for plasmas with current up to 300 kA and flat-top duration up to 4s. Keywords: FTU, Liquid lithium limiter, L-H transition, X-point, Plasma elongation

Home care is greatly expanding because of the savings it offers by avoiding unnecessary hospitalization and also because patients benefit from being in their own home environments. Since 1988, Societa Assistenza Malato Oncologico Terminale (SAMOT) has organized a pain relief and home palliative care unit for terminal cancer patients. Objectives, difficulties, protocols, and achievements of 4 years of experience were examined, and the findings of the various years were compared. Our results suggest that considerable progress has been made in home palliative care. There are still social and cultural difficulties to overcome, however.