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Thin tungsten coatings on Inconel superalloy were prepared in the frame of the COMPASS Upgrade project, the new high magnetic field tokamak with metallic walls now under construction at the Institute of Plasma Physics in Prague. The coatings will be used to cover the in-vessel Inconel tiles which are not in direct contact with the plasma, representing ∼75 % of the total first wall surface. This paper describes the latest developments in the coating technology by three different suppliers and presents a comprehensive comparative characterization of the different coatings (morphology, thickness, adhesion) produced by physical vapor deposition. Coatings with thicknesses in the range 7–20 µm were successfully produced and analyzed. Overall, they all show no blisters, no delamination nor other defects such as cracks, large pores, inclusions or droplets, and a good adhesion. All the coatings featured sufficient purity and homogeneity in thickness complying with COMPASS Upgrade specifications. The qualification of the coatings under high heat flux tests is presented in a companion paper.

Coating Inconel tiles by tungsten is a necessary step towards the full tungsten first wall coverage of the COMPASS-U tokamak. Thin tungsten coatings on Inconel based on physical vapor deposition were successfully produced consequently to a programme of R&D within the COMPASS Upgrade project by three different suppliers. This contribution presents the qualification phase of these tungsten coatings under COMPASS-U relevant high heat fluxes (50 cycles at 10 MW/m2, 100 cycles at 30 MW/m2) in the neutral beam test facility GLADIS. The behavior of the different tungsten coatings during exposures and the main conclusions from the post-mortem analysis are presented. The most important result is that under COMPASS-U relevant heat fluxes, no damage was observed on the front face of all samples, proving the practicability of such coatings for fusion application.

Plasma-facing materials in tokamak nuclear reactors are subject to intense heat and radiation. The design of liquid metal (LM) divertor walls has several advantages over the solid counterparts. It is important to understand the migration and re-deposition of LMs such as lithium (Li), lead (Pb) or even tin (Sn). This work presents CF LIBS depth profile analysis of the deposited layer on Ni–Cr based screws situated at different places in the COMPASS tokamak vessel during the Capillary Porous System (CPS) LM (LiSn) divertor testing campaign. The experiment was carried out at low Ar pressure (100 mbar) using ns Nd:YAG laser operating at 532 nm. The spectra were recorded using echelle spectrometer (M5000, Andor Tech.) equipped with an iCCD camera (iStar DH743, Andor Tech.). The quantification was based on observed C I, B I, Li I and (weak) Sn I lines, in addition to trace elements (Ca, Na, K) and main matrix elements (Cr, Ni, Fe). The evaluation of electron temperature and density was based on Cr I-II, Ni-II and Fe I-II lines. The LM based content is estimated to be Li (0.3-0.81 at.%) and Sn (0.1-0.23 at.%) with Li/Sn at.% ratio ranging from 3 to 3.5 (wt.% ratio of 0.17-0.20). In practise, the limit of detection of Sn by CF LIBS for this type of Cr-Ni-Fe based matrix is observed to be ∼0.1 at.%.

The consequences of tungsten (W) cracking on divertor lifetime and plasma operation are high priority issues for ITER. One actively cooled ITER-like plasma facing unit (PFU) has been pre-damaged in a High Heat Flux (HHF) facility before its installation in WEST in order to assess the damage evolution after tokamak plasma exposure. The resulting pre-damage exhibits micrometer-size crack network and high roughness on the tungsten monoblock (MB) top surface. A total of 10 MBs, equally distributed on the low and high field sides of the lower divertor, have been pre-damaged among the 35 radially aligned MBs characteristic of the WEST PFU. Subsequent plasma exposure was carried out, from the first breakdown achieved in WEST (in 2017) until the removal of the damaged PFU three years later (2020). On top of the whole WEST plasma exposure (covering C1-C4 experimental campaigns), a dedicated experiment has also been performed in the frame of the EU work program to maximize the power and energy loads on one of the damaged MBs featuring a “crack network” pattern. The MB top surface, including both “crack network” damage and “healthy” (undamaged) areas, was monitored with a high spatial resolution IR camera to detect any potential evolution of the damage pulse after pulse. This paper describes the full plasma exposure achieved in the WEST tokamak (including large number of steady-state and transient heat loading cycles), the dedicated “damaged PFU exposure” experiment together with the experimental results (heat loading on the damaged MBs). Post-mortem measurement reveals significant broadening of the cracks and new cracks in the electron beam loaded area only.