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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Invited Review Paper -R2 CEI SUN 3 0 O S T ITritium retention inside the vacuum vessel has emerged as a potentially serious constraint in the operation of the International Thermonuclear Experimental Reactor (ITER). In this paper we review recent tokamak and laboratory data on hydrogen, deuterium and tritium retention for materials and conditions which are of direct relevance to the design of ITER. These data, together with significant advances in understanding the underlying physics, provide the basis for modelling predictions of the tritium inventory in ITER.We present the derivation, and discuss the results, of current predictions both in terms of implantation and codeposition rates, and critically discuss their uncertainties and sensitivity to important design and operation parameters such as the plasma edge conditions, the surface temperature, the presence of mixed-materials, etc. These analyses are consistent with recent tokamak findings and show that codeposition of tritium occurs on the divertor surfaces primarily with carbon eroded from a limited area of the divertor near the strike zones. This issue remains an area of serious concern for ITER. The calculated codeposition rates for ITER are relatively high and the in-vessel tritium inventory limit could be reached, under worst assumptions, in approximately a week of continuous operation.We discuss the implications of these estimates on the design, operation and safety of ITER and present a strategy for resolving the issues. We conclude that as long as carbon is used in ITER, the efficient control and removal of the codeposited tritium is essential. There is a critical need to develop and test in-situ cleaning techniques and procedures that are beyond the current experience of present-day tokamaks. We review some of the principal methods that are being investigated and tested, in conjunction with the R&D work still required to extrapolate their applicability to ITER. Finally, unresolved issues are identified and recommendations are made on potential R&D avenues for their resolution.

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This study focuses on the removal of trapped D from thick codeposits on JET divertor tiles via thermo-oxidation. The tiles were removed from the JET Mark II Gas Box divertor after the 1998–2001 campaign. These codeposits have Be concentrations of up to ∼60% Be/(Be + C) and their thicknesses range from 10 to 270 µm. Laser thermal desorption spectroscopy was used to determine the D removal rates and final remaining D concentrations following oxidation. Estimates of the carbon removed during oxidation were obtained from mass-loss measurements. The initial rate of D removal was found to be much higher for the thick codeposits of this study than for the previously studied codeposits with thicknesses in the range 1–5 µm (from TFTR, DIII-D and JET). This is despite the large Be concentrations. For oxidation performed at 623 K (350 °C) and 21 kPa (160 Torr) O2 pressure the initial D removal rates were found to increase linearly with increasing ‘inherent’ D content; about 50% of the inherent D was removed from all specimens in the first 15 min—independent of Be content and codeposit thickness. Following 8 h of oxidation, the fraction of D removed was >85% for all specimens, again, independent of Be content and thickness.

The similar poloidal patterns of divertor erosion and redeposition of the ASDEX Upgrade, DIII-D and JET tokamaks are examined by comparing hydrogen isotope retention at the surfaces of divertor tiles. The outer divertor strike point undergoes net erosion. The inner divertor is a region of net redeposition, with deuterium co-deposition rates ∼20 μg m2 s-1 for all three devices; despite their differences in geometry, size and plasma facing materials. This extrapolates to an unacceptably high tritium retention rate in a DT burning steady state tokamak.