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The evaluation of the impact of plasma-facing components damage on subsequent plasma operation is an important issue for ITER. During the first phase of the operation of WEST, a few ITER-like divertor plasma-facing units (PFUs) have been installed on the lower divertor. One PFU was pre-damaged under electron beam gun thermal loading, before its installation in WEST, and the subsequent evolution of the damage was studied after the WEST plasma exposure. This paper presents the procedure followed to get the pre-damaged PFU. It consists of the characterization of the response of tungsten samples representative of WEST PFU under high heat flux (HHF) loading, the selection of damage (namely, small cracks, crack network, crack network, and W melt droplets). Finally, according to the WEST plasma loading conditions, the blocks with damage within the PFU and the position of the pre-damaged PFU on the WEST lower divertor are attributed. The first results obtained after an initial plasma exposure in WEST lead to the assessment, as expected with regard to the heat loading conditions, that no major surface aspect modification was found. This result emphasized the possibility of implementing as pre-damaged small droplets of melted tungsten in a high heat-loaded zone for a future WEST experimental campaign.

Liquid metal divertors aim to provide a more robust alternative to conventional tungsten divertors. However, they still require a solid substrate to confine the liquid metal. This work proposes a novel design philosophy for liquid metal divertor targets, which allows for a two orders of magnitude reduction of thermal stresses compared to the state-of-the-art monoblock designs. The main principle is based on a 3D-printed tungsten structure, which has low connectedness in the direction perpendicular to the thermal gradient, and as a result also short length scales. This allows for thermal expansion. Voids in the structure are filled with liquid lithium which can conduct heat and reduce the surface temperature via vapor shielding, further suppressing thermal stresses. To demonstrate the effectiveness of this design strategy, an existing liquid metal concept is re-designed, fabricated, and tested on the linear plasma device Magnum-PSI. The thermo-mechanical finite element method analysis of the improved design matches the temperature response during the experiments, and indicates that thermal stresses are two orders of magnitude lower than in the conventional monoblock designs. The relaxation of the strength requirement allows for much larger failure margins and consequently for many new design possibilities.

Six tungsten grades were irradiated in the Belgian material test reactor (BR2) and characterized by Vickers hardness tests in order to investigate the irradiation-induced hardening. These tungsten grades included: Plansee (Austria) ITER specification tungsten, ALMT (Japan) ITER specification tungsten, two products from KIT (Germany) produced by powder injection molding (PIM) and strengthened by 1% TiC and 2% Y2O3 dispersed particles, and rolled tungsten strengthened by 0.5% ZrC from ISSP (China). The materials were irradiated face-to-face at three temperatures equal to 600 °C, 1000 °C, and 1200 °C to the dose of ∼1 dpa. The Vickers hardness tests under 200 gf (HV0.2) were performed at room temperature. The Vickers hardness increases as the irradiation temperature increases from 600 to 1000 °C for all materials, except for the ZrC-reinforced tungsten, for which the increase of hardness does not depend on irradiation temperature. The irradiation-induced hardness decreases after irradiation at 1200 °C. This is a result of defect annealing enhanced by thermally activated diffusion. However, even at 1200 °C, the impact of neutron irradiation on the hardness increase remains significant; the hardness increases by ∼30 to 60% compared to the non-irradiated value. In the case of TiC-strengthened material, the irradiation hardening progressively raises with irradiation temperature, which cannot be explained by the accumulation of neutron irradiation defects solely.

With the aim of a possible improvement of the material specification for tungsten, five different tungsten products by different companies and by different production technologies (forging and rolling) are subject to a materials characterization program. Tungsten produced by forging results in an uniaxial elongated grain shape while rolled products have a plate like grain shape which has an influence on the mechanical properties of the material. The materials were investigated with respect to the following parameters: hardness measurements, microstructural investigations, tensile tests and recrystallisation sensitivity tests at 3 different temperatures. The obtained results show that different production processes have an influence on the resulting anisotropic microstructure and the related material properties of tungsten in the as-received state. Additionally, the recrystallization sensitivity varies between the different products, what could be a result of the different production processes. Additionally, two tungsten products were exposed to thermal shocks. The obtained results show that the improved recrystallisation behaviour has no major impact on the thermal shock performance.

Experiments were performed in three different facilities in order to investigate the impact of combined steady state deuterium plasma exposure and ELM-like thermal shock events on the performance of ultra high purity tungsten. The electron beam facility JUDITH 1 was used to simulate pure thermal loads. In addition the linear plasma devices PSI-2 and Pilot-PSI have been used for successive as well as simultaneous exposure where the transient heat loads were applied by a high energy laser and the pulsed plasma operation, respectively. The results show that the damage behaviour strongly depends on the loading conditions and the sequence of the particle and heat flux exposure. This is due to hydrogen embrittlement and/or a higher defect concentration in the tungsten near surface region due to supersaturation of hydrogen. The different results in terms of damage formation from both linear plasma devices indicate that also the plasma parameters such as particle energy, flux and fluence, plasma impurities and the pulse shape have a strong influence on the damage performance. In addition, the different loading methods such as the scanning with the electron beam in contrast to the homogeneous exposure by the laser leads to an faster increase of the surface roughness due to plastic deformation.

In the experimental fusion reactor ITER, beryllium will be applied as first wall armor material. However, the ITER-like wall project at JET already experienced that the relatively low melting temperature of beryllium can easily be exceeded during plasma operation. Therefore, a detailed study was carried out on S-65 beryllium under various transient, ITER-relevant heat loads that were simulated in the electron beam facility JUDITH 1. Hereby, the absorbed power densities were in the range of 0.15–1.0 GW m−2 in combination with pulse durations of 1–10 ms and pulse numbers of 1–1000. In metallographic cross sections, the emergence of a transition region in a depth of ~70–120 µm was revealed. This transition region was characterized by a strong segregation of oxygen at the grain boundaries, determined with energy dispersive x-ray spectroscopy element mappings. The oxide segregation strongly depended on the maximum temperature reached at the end of the transient heat pulse in combination with the pulse duration. A threshold for this process was found at 936 °C for a pulse duration of 10 ms. Further transient heat pulses applied to specimens that had already formed this transition region resulted in the overheating and melting of the material. The latter occurred between the surface and the transition region and was associated with a strong decrease of the thermal conductivity due to the weakly bound grains across the transition region. Additionally, the transition region caused a partial separation of the melt layer from the bulk material, which could ultimately result in a full detachment of the solidified beryllium layers from the bulk armor. Furthermore, solidified beryllium filaments evolved in several locations of the loaded area and are related to the thermally induced crack formation. However, these filaments are not expected to account for an increase of the beryllium net erosion.

Surface damage and structure evolution of the full tungsten ITER divertor under transient heat loads is a key concern for component lifetime and plasma operations. Recrystallization caused by transients and steady-state heat loads can lead to degradation of the material properties and is therefore one of the most serious issues for tungsten armor. In order to investigate the thermal response of the recrystallized tungsten under edge localized mode-like transient thermal loads, fully recrystallized tungsten samples with different average grain sizes are exposed to cyclic thermal shocks in the electron beam facility JUDITH 1. The results indicate that not only does the microstructure change due to recrystallization, but that the surface residual stress induced by mechanical polishing strongly influences the surface cracking behavior. The stress-free surface prepared by electro-polishing is shown to be more resistant to cracking than the mechanically polished one. The resulting surface roughness depends largely on the loading conditions instead of the recrystallized-grain size. As the base temperature increases from room temperature to 400 °C, surface roughening mainly due to the shear bands in each grain becomes more pronounced, and sub-grains (up to 3 μm) are simultaneously formed in the sub-surface. The directions of the shear bands exhibit strong grain-orientation dependence, and they are generally aligned with the traces of {1 1 2} twin habit planes. The results suggest that twinning deformation and dynamic recrystallization represent the predominant mechanism for surface roughening and related microstructure evolution.

Plasma facing components in future thermonuclear fusion devices will be subjected to intense transient thermal loads due to type I edge localized modes (ELMs), plasma disruptions, etc. To exclude irreversible damage to the divertor targets, local energy deposition must remain below the damage threshold for the selected wall materials. For monolithic tungsten (pure tungsten and tungsten alloys) power densities above ≈0.3 GW m−2 with 1 ms duration result in the formation of a dense crack network. Thin tungsten coatings for the so-called ITER-like wall in JET, which have been deposited on a two-directional carbon–fibre composite (CFC) material, are even less resistant to thermal shock damage; here the threshold values are by a factor of 2 lower. First ELM-simulation experiments with high cycle numbers up to 104 cycles on actively cooled bulk tungsten targets do not reveal any cracks for absorbed power densities up to 0.2 GW m−2 and ELM-durations in the sub-millisecond range (0.8 ms); at somewhat higher power densities (0.27 GW m−2, Δt = 0.5 ms) cracks have been detected for 106 cycles.

The high-flux deuterium plasma impinging on a divertor degrades the long-term thermo-mechanical performance of its tungsten plasma-facing components. A prime actor in this is hydrogen embrittlement, a degradation phenomenon that involves the interactions between hydrogen and dislocations, the primary carriers of plasticity. Measuring such nanoscale interactions is still very challenging, which limits our understanding. Here, we demonstrate an experimental approach that combines thermal desorption spectroscopy (TDS) and nanoindentation, allowing to investigate the effect of hydrogen on the dislocation mobility in tungsten. Dislocation mobility was found to be reduced after deuterium injection, which is manifested as a 'pop-in' in the indentation stress-strain curve, with an average activation stress for dislocation mobility that was more than doubled. All experimental results can be confidently explained, in conjunction with experimental and numerical literature findings, by the simultaneous activation of three mechanisms responsible for dislocation pinning: (i) hydrogen trapping at pre-existing dislocations, (ii) hydrogen-induced vacancies, and (iii) stabilization of vacancies by hydrogen, contributing respectively 38%, 52%, and 34% to the extra activation stress. These mechanisms are considered to be essential for the proper understanding and modeling of hydrogen embrittlement in tungsten.