Recently the design of the divertor tokamak test (DTT) Facility divertor has been modified and consolidated. The new divertor design presents significant geometrical differences compared to the previous ITER-like one, including the presence of a more flattened dome shape. This paper presents a complete 3D numerical analysis of the neutral gas dynamics inside the DTT subdivertor area for the latest divertor design. The analysis has been performed based on the direct simulation Monte Carlo method by applying the DIVGAS simulator code. SOLEDGE2D-EIRENE plasma simulations have been performed for a deuterium plasma scenario at the maximum additional power in partially detached condition achieved by neon impurity seeding and the extracted information about the neutral particles has been imposed as incoming boundary conditions. The pumping efficiency of the DTT divertor is examined by considering various cases with respect to the pumping probability and the effect of the toroidal and poloidal leakages is quantified. The results show that a significant percentage of the incoming flux of neutrals returns back to the plasma site through the entry gaps (60% for deuterium and 40% for neon), and, consequentially, only a small percentage (~2%-15%) of the incoming flux can be pumped out from the system. The toroidal leakages affect significantly the pumping performance of the divertor causing a significant decrease in the pumped flux (and also in the pressure at the pumping opening) about 37%-47% and 43%-56% for deuterium and neon respectively. It is discussed how many pumping ports are needed depending on the achievable pumping performance per port. The number can be reduced by closing the toroidal gaps. The analysis shows that enlarging the poloidal gaps by a factor of two causes a significant increase in the poloidal flux losses by a factor 1.7. It is also illustrated how the presence of the cooling pipes leads to conductance losses.

The provision of a particle and power exhaust solution which is compatible with first-wall components and edge-plasma conditions is a key area of present-day fusion research and mandatory for a successful operation of ITER and DEMO. The work package plasma-facing components (WP PFC) within the European fusion programme complements with laboratory experiments, i.e. in linear plasma devices, electron and ion beam loading facilities, the studies performed in toroidally confined magnetic devices, such as JET, ASDEX Upgrade, WEST etc. The connection of both groups is done via common physics and engineering studies, including the qualification and specification of plasma-facing components, and by modelling codes that simulate edge-plasma conditions and the plasma-material interaction as well as the study of fundamental processes. WP PFC addresses these critical points in order to ensure reliable and efficient use of conventional, solid PFCs in ITER (Be and W) and DEMO (W and steel) with respect to heat-load capabilities (transient and steady-state heat and particle loads), lifetime estimates (erosion, material mixing and surface morphology), and safety aspects (fuel retention, fuel removal, material migration and dust formation) particularly for quasi-steady-state conditions. Alternative scenarios and concepts (liquid Sn or Li as PFCs) for DEMO are developed and tested in the event that the conventional solution turns out to not be functional. Here, we present an overview of the activities with an emphasis on a few key results: (i) the observed synergistic effects in particle and heat loading of ITER-grade W with the available set of exposition devices on material properties such as roughness, ductility and microstructure; (ii) the progress in understanding of fuel retention, diffusion and outgassing in different W-based materials, including the impact of damage and impurities like N; and (iii), the preferential sputtering of Fe in EUROFER steel providing an in situ W surface and a potential first-wall solution for DEMO.