In fusion reactors, surfaces of plasma facing components (PFCs) are exposed to high heat and particle flux. Tungsten and Copper alloys are primary candidates for plasma facing materials (PFMs) and coolant tube materials, respectively, mainly due to high thermal conductivity and, in the case of tungsten, its high melting point. In this paper, recent understandings and future issues on responses of tungsten and Cu alloys to fusion environments (high particle flux (including T and He), high heat flux, and high neutron doses) are reviewed. This review paper includes; Tritium retention in tungsten (K. Schmid and M. Balden), Impact of stationary and transient heat loads on tungsten (J.W. Coenen and Th. Loewenhoff), Helium effects on surface morphology of tungsten (Y. Ueda and A. Ito), Neutron radiation effects in tungsten (A. Hasegawa), and Copper and copper alloys development for high heat flux components (C. Hardie, M. Porton, and M. Gilbert).

The effects of ion irradiation in materials for research are usually limited to a shallow surface layer of the order of one micrometre in depth. Determining the mechanical properties of such irradiated materials requires techniques with high spatial resolution. Nanoindentation is a relatively simple method for investigating these shallow layers with the advantage that statistically rich data sets for elastic and plastic property values can be generated. However, interpretation of the results requires and understanding of the material response, including the extent of the plastic zone with respect to the irradiated layer, pile-up or sink-in of material around the indentation site that affect the calculated contact area and hence derived mechanical property values. An Fe + self-irradiated Fe12%Cr alloy was investigated with three different indenter tip geometries, a Cube corner, Berkovich and 10 micron radius indenter. Sharp indenters provide sufficiently small plastic zones to be contained within the irradiated layer but pop-in events and pile-up need to be taken into account for correct interpretation of the mechanical properties of the irradiated material.

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Results of the first dust survey in JET with the ITER-Like Wall (JET-ILW) are presented. The sampling was performed using adhesive stickers from the divertor tiles where the greatest material deposition was detected after the first JET-ILW campaign in 2011–2012. The emphasis was especially on sampling and analysis of metal particles (Be and W) with the aim to determine the composition, size, surface topography and internal dust structure using a large set of methods: high-resolution scanning and transmission electron microscopy, focused ion beam, electron diffraction and also wavelength and energy dispersive x-ray spectroscopy. The most important was the identification of beryllium dust both in the form of flakes and droplets with dimensions in the micrometer range. Tungsten, molybdenum, inconel constituents were identified along with many impurity species. The particles are categorised and the origin of the various constituents discussed.

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The ITER project requires additional heating by two neutral beam injectors, each accelerating to 1 MV a 40 A beam of negative deuterium ions, to deliver to the plasma a power of about 17 MW for one hour. As these requirements have never been experimentally met, it was Nuclear Fusion Progress in the realization of the PRIMA neutral beam test facility

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