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.

Trapping of tritium at lattice damage from fusion neutron irradiation is expected to increase the tritium inventory in tungsten components in ITER. The magnitude of this increase depends on the concentration of traps that are produced, and on the depth to which the increased tritium retention extends into the material. Experiments to address these issues are described, in which displacement damage by ion irradiation was used as a surrogate for neutron damage. Irradiated samples were exposed to high flux deuterium plasma to simulate divertor conditions. The resulting deuterium content was measured by nuclear reaction analysis. Measurements were done at various damage levels up to those expected from the end-of-life neutron fluence in ITER. These experiments determine the number of traps produced by displacement damage and the rate at which they are filled during exposure to plasma. The role of defect annealing was explored through plasma exposures at various temperatures. In addition to trapping at damage, near-surface retention from internal precipitation was observed at lower temperatures. Addition of 5% helium to the deuterium plasma greatly reduced D retention by precipitation by localizing it closer to the surface. Results from these experiments indicate that the contribution to tritium inventory in ITER from trapping at neutron damage should be small.

National Spherical Torus Experiment [which M. Ono et al., Nucl. Fusion 40, 557 (2000)] high-power divertor plasma experiments have shown, for the first time, that benefits from lithium coatings applied to plasma facing components found previously in limited plasmas can occur also in high-power diverted configurations. Lithium coatings were applied with pellets injected into helium discharges, and also with an oven that directed a collimated stream of lithium vapor toward the graphite tiles of the lower center stack and divertor. Lithium oven depositions from a few milligrams to 1g have been applied between discharges. Benefits from the lithium coatings were sometimes, but not always, seen. These benefits sometimes included decreases in plasma density, inductive flux consumption, and edge-localized mode occurrence, and increases in electron temperature, ion temperature, energy confinement, and periods of edge and magnetohydrodynamic quiescence. In addition, reductions in lower divertor D, C, and O luminosity were measured.

W targets are exposed at fixed temperature in the range ∼420–1100 K, to either pure D2, D2–δHe (0.1 < δ < 0.25), or D2–δHe–γAr (γ = 0.03) mixture plasma, or He pretreatment plasma followed by exposure to D2 plasma. A strong reduction in D retention is found for exposure temperature above 450 K and incident He-ion fluence exceeding ∼1024 m−2. Reduced D retention values lie well below that measured on D2 plasma-exposed reference targets, and the scatter in retention values reported in the literature. A small level of Ar admixture to D2–0.1He plasma, leading to an Ar ion density fraction of ∼3%, is found to have minimal effect on the D inventory reduction caused by He. In targets with reduced inventory, nuclear-reaction analysis reveals shallow D trapping (<50 nm), in the same locale as nanometre-sized bubbles observed using transmission electron microscopy. It is suggested that near-surface bubbles grow and interconnect, forming pathways leading back to the plasma–material interaction surface, thereby interrupting transport to the bulk and reducing D retention.

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