The major increase in discharge duration and plasma energy in a next step DT fusion reactor will give rise to important plasma-material effects that will critically influence its operation, safety and performance. Erosion will increase to a scale of several centimetres from being barely measurable at a micron scale in today's tokamaks. Tritium co-deposited with carbon will strongly affect the operation of machines with carbon plasma facing components. Controlling plasma-wall interactions is critical to achieving high performance in present day tokamaks, and this is likely to continue to be the case in the approach to practical fusion reactors. Recognition of the important consequences of these phenomena stimulated an internationally co-ordinated effort in the field of plasma-surface interactions supporting the Engineering Design Activities of the International Thermonuclear Experimental Reactor project (ITER), and significant progress has been made in better understanding these issues. The paper reviews the underlying physical processes and the existing experimental database of plasma-material interactions both in tokamaks and laboratory simulation facilities for conditions of direct relevance to next step fusion reactors. Two main topical groups of interaction are considered: (i) erosion/redeposition from plasma sputtering and disruptions, including dust and flake generation and (ii) tritium retention and removal. The use of modelling tools to interpret the experimental results and make projections for conditions expected in future devices is explained. Outstanding technical issues and specific recommendations on potential R&D avenues for their resolution are presented.

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Plasma{material interactions in current tokamaks and their implications for next step fusion reactors

Hydrogen isotope retention and recycling in fusion reactor plasma-facing components

Development of tungsten armor and bonding to copper for plasma-interactive components

Deuterium retention in tungsten in dependence of the surface conditions

Understanding the interaction between atomic hydrogen and solid tungsten is important for the development of fusion reactors in which proposed tungsten walls would be bombarded with high energy particles including hydrogen isotopes. Here, we report results from periodic density-functional theory calculations for three crucial aspects of this interaction: surface-to-subsurface diffusion of H into W, trapping of H at vacancies, and H-enhanced decohesion, with a view to assess the likely extent of hydrogen isotope incorporation into tungsten reactor walls. We find energy barriers of (at least) 2.08 eV and 1.77 eV for H uptake (inward diffusion) into W(001) and W(110) surfaces, respectively, along with very small barriers for the reverse process (outward diffusion). Although H dissolution in defect-free bulk W is predicted to be endothermic, vacancies in bulk W are predicted to exothermically trap multiple H atoms. Furthermore, adsorbed hydrogen is predicted to greatly stabilize W surfaces such that decohesion (fracture) may result from high local H concentrations.

https://www.cambridge.org/core/services/aop-cambridge-core/content/view/S0884291400006075

Hydrogen in tungsten: Absorption, diffusion, vacancy trapping, and decohesion

Deuterium permeation studies for polycrystalline tungsten foil have been conducted to provide data for estimating tritium transport and trapping in tungsten-clad divertors proposed for advanced fus...

R.J. Pawelko

Papers

Deuterium Transport and Trapping in Polycrystalline Tungstena

Deuterium/tritium behavior in Flibe and Flibe-facing materials

Steam-chemical reactivity for irradiated beryllium

Hydrogen transport behavior of beryllium

Hydrogen permeation properties of plasma-sprayed tungsten☆