• Energy Research

ITER will use beryllium as a plasma-facing material in the main chamber, covering a total surface area of about 620 m2. Given the importance of beryllium erosion and co-deposition for tritium retention in ITER, significant efforts have been made to understand the behaviour of beryllium under fusion-relevant conditions with high particle and heat loads. This paper provides a comprehensive report on the state of knowledge of beryllium behaviour under fusion-relevant conditions: the erosion mechanisms and their consequences, beryllium migration in JET, fuel retention and dust generation. The paper reviews basic laboratory studies, advanced computer simulations and experience from laboratory plasma experiments in linear simulators of plasma–wall interactions and in controlled fusion devices using beryllium plasma-facing components. A critical assessment of analytical methods and simulation codes used in beryllium studies is given. The overall objective is to review the existing set of data with a broad literature survey and to identify gaps and research needs to broaden the database for ITER.

ITER will use beryllium as a plasma-facing material in the main chamber, covering a total surface area of about 620 m2. Given the importance of beryllium erosion and co-deposition for tritium retention in ITER, significant efforts have been made to understand the behaviour of beryllium under fusion-relevant conditions with high particle and heat loads. This paper provides a comprehensive report on the state of knowledge of beryllium behaviour under fusion-relevant conditions: the erosion mechanisms and their consequences, beryllium migration in JET, fuel retention and dust generation. The paper reviews basic laboratory studies, advanced computer simulations and experience from laboratory plasma experiments in linear simulators of plasma–wall interactions and in controlled fusion devices using beryllium plasma-facing components. A critical assessment of analytical methods and simulation codes used in beryllium studies is given. The overall objective is to review the existing set of data with a broad literature survey and to identify gaps and research needs to broaden the database for ITER.

Post mortem analyses of dust collected in Alcator C-Mod have highlighted a production of large size dust particles. The quantities of such large particles are higher than in any other tokamak. They are divided in two classes as a function of their shape and consequently, their origin. Rounded dust particles such as spheres and splashes constitute the first class. These particles are the result of high heat loads on various leading edges of plasma facing components and possibly, their melting during plasma operation. The heated or already molten material can be destabilized during disruptions and droplets are emitted across the vacuum chamber. After solidification, the resulting rounded particles are either in pure elements or in alloys. Flake-like dust particles, which are mainly due to light material coating delamination, constitute the second class of dust particles.

Post mortem analyses of dust collected in Alcator C-Mod have highlighted a production of large size dust particles. The quantities of such large particles are higher than in any other tokamak. They are divided in two classes as a function of their shape and consequently, their origin. Rounded dust particles such as spheres and splashes constitute the first class. These particles are the result of high heat loads on various leading edges of plasma facing components and possibly, their melting during plasma operation. The heated or already molten material can be destabilized during disruptions and droplets are emitted across the vacuum chamber. After solidification, the resulting rounded particles are either in pure elements or in alloys. Flake-like dust particles, which are mainly due to light material coating delamination, constitute the second class of dust particles.

A novel identification technique of hydrogen transport parameters using FESTIM (Finite Element Simulation of Tritium In Materials) has been demonstrated. FESTIM is a finite element code developed with FEniCS performing hydrogen transport simulations. The trapping parameters (detrapping energies and trap densities) are identified for various materials (Tungsten, Aluminium, EUROFER and Beryllium) by automatically reproducing thermo-desorption experiments. Several optimisation algorithms are tested and the Nelder–Mead algorithm shows the best efficiency. An optimisation test problem with five free parameters took only a few hours to solve whereas optimisation cases with two free parameters took a few minutes. Limitations of this technique are shown and discussed.