Current R&D activities on materials for fusion power reactors are mainly focused on plasma facing, structural and tritium breeding materials for plasma facing (first wall, divertor) and breeding blanket components. Most of these activities are being performed in Europe, Japan, the People's Republic of China, Russia and the USA. They relate to the development of new high temperature, radiation resistant materials, the development of coatings that will act as erosion, corrosion, permeation and/or electrical/MHD barriers, characterization of candidate materials in terms of mechanical and physical properties, assessment of irradiation effects, compatibility experiments, development of reliable joints, and development and/or validation of design rules. Priorities defined worldwide in the field of materials for fusion power reactors are summarized, as well as the main achievements obtained during the last few years and the near-term perspectives in the different investigation areas.

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Status of R&D activities on materials for fusion power reactors

Thermonuclear fusion is a potential source of cleaner and safer energy for the future. Its technological realization depends on the development of materials able to survive and function in extreme conditions. This article reviews the applications of thermally sprayed coatings for fusion reactor materials. First, the principle and purpose of fusion is briefly introduced, and technological objectives are mentioned. Material-environment interactions are summarized, together with materials requirements and the role of coatings. Then, specific applications—e.g., the plasma facing components—are reviewed, focusing on application issues as well as issues related to thermal spray processing and specific properties of the respective materials. An overview of specific materials testing methods is also provided.

Thermal Spray Coatings for Fusion Applications—Review

https://iopscience.iop.org/article/10.1088/1741-4326/aa7e4e/pdf

Functional materials for breeding blankets—status and developments

The study of melting behaviour and crack formation, which occurs during short transient events, is of significant importance for the qualification of materials for future fusion devices. Heat load simulations at room temperature with a pulse duration of 5 ms have been performed on beryllium (S65C) and various tungsten grades with unidirectionally elongated grains, ITER candidate materials for the first wall (Be) and the divertor region (W), at several power loads below and above the melting threshold (∼50 MW m −2 ·s 1/2 for W, ∼28 MW m −2  s 1/2 for Be). Crack formation and surface roughening have been studied at single and in case of tungsten also multiple shots. Significant differences in the crack resistance and the crack pattern of the various tungsten grades below the melting threshold have been determined and further material degradation has been found after multiple shots. In microstructural and metallographic studies, the material damage has been qualified and quantified. Doing so the material's ability to avoid premature failure as well as plasma contamination (by evaporating tungsten particles below the melting threshold) during ITER operation has been characterised.

F. Scaffidi-Argentina

Papers

Beryllium R&D for fusion applications

The status of beryllium technology for fusion

Beryllium for fusion application – recent results