The main differences between a future power reactor and ITER are the much longer operation time (e. g., >108s vs. >107s), high duty cycle, and the higher temperature of the fluid to cool the PFCs to maintain a high plant energy conversion efficiency. The higher surface temperature of the PFC will affect particle recycling, tritium uptake, chemical erosion, and material-mixing effects.

Erosion rates at the divertor target are very diffi­cult to predict in conventional fusion power plant concepts with solid high-Z targets because the net erosion or deposition is strongly dependent on plasma parameters. The fraction of ions arriving above the sputtering threshold is crucial, as is the efficiency of the prompt local redeposition. ELMs have not really been considered in this context but we can see from the analysis in Section 4.19.6.2.2.2 that ELMs in power plant systems will have to be extremely small — much smaller than will be allow­able in ITER, which still has a relatively low duty cycle. The ideal would in fact be a quiescent ELM free high density steady state edge plasma.

Calculations of the minimum erosion rate for the main wall are somewhat more robust as there has to be a hot plasma in the main chamber and the rate of leakage of neutrals into the main chamber from the divertor can be calculated using Monte-Carlo codes. In a recent study, Be, C, Fe, Cu, Mo, and W walls were compared,206 with the conclusion that in all cases the erosion rate was 1-2 t per year of continuous operation. A Be or CFC wall will erode too rapidly in a reactor and the large amount of eroded material might give rise to deleterious problems as far as control of the tritium and dust inventories are concerned. A medium-Z material, such as Fe, does not seem to be acceptable purely from the standpoint of erosion lifetime. As molybdenum is unfavorable for long-term activation problems, W is the best and only solution we have available for a reactor. Effects of plasma contamination from Mo and W at the wall of tokamaks are being addressed in Alcator C-Mod, ASDEX-Upgrade and in the near-future at JET.

There is considerable gross erosion by sputtering for all materials. The contributions ofions and neutrals from the plasma to this erosion are ofthe same order of magnitude. The integrated total erosion due to ions and the energetic neutrals for the different wall mate­rials (Figure 24) show that because of the larger sput­tering yields for the low-Z materials, the number of atoms eroded for these materials is a factor of 10-20 larger than for high-Z materials such as W. However, the total mass loss is similar for all materials, up to several kilograms per day or about 1 t per year.

The maximum wall thinning for the low-Z mate­rials is about 3.5 mmyear~ , while for high-Z materi­als, such as W, it is 0.22 mm year, that is, lower by about a factor of 15. These values are in reasonable agreement with erosion measurements at the JET vessel walls.223 With respect to wall thinning, W is favorable for the use at the vessel walls because it has the longest ‘erosion lifetime’ (Figure 24(b)). With respect to plasma contamination, the probability of the eroded atoms entering into the plasma core, their lifetime in the plasma core, and the tolerable concen­tration of these ions in a burning fusion plasma all have to be taken into account.206 The tolerable con­centration of W in the plasma is nearly three orders of magnitude lower than for low-Z atoms, such as Be and C. However, recent observations have shown that W can be effectively removed from the plasma center by central heating.224 As this central heating is natural for burning plasmas, W may be a possible plasma­facing material, even from the viewpoint of plasma contamination. The ion and neutral flux densities on the vessel walls are of the order of 102°m~2s~1, which may be critical with respect to the tritium implantation, accumulation in and permeation through the vessel walls.