Abstract:Plasma-wall interaction (PWI) is important for the material choice in ITER and for the plasma scenarios compatible with material constraints. In this paper different aspects of the PWI are assessed in their importance for the initial wall materials choice: CFC for the strikepoint tiles, W in the divertor and baffle and Be on the first wall. Further material options are addressed for comparison, such as W divertor / Be first wall and all-W or all-C.One main parameter in this evaluation is the particle flux to the main vessel wall. One detailed plasma scenario exists for a Q=10 ITER discharge [ 1 ] which was taken as the basis of further erosion and tritium retention evaluations. As the assessment of steady state wall fluxes from a scaling of present fusion devices indicates that global wall fluxes may be a factor of 4±3 higher, this margin has been adopted as uncertainty of the scaling.With these wall and divertor fluxes, important PWI processes such as erosion and tritium accumulation have been evaluated:• It was found that the steady state erosion is no problem for the lifetime of plasma-facing divertor components. Be wall erosion may pose a problem in case of a concentration of the wall fluxes to small wall areas. ELM erosion may drastically limit the PFC lifetime if ELMs are not mitigated to energies below 0.5 MJ.• Dust generation is still a process which requires more attention. Conversion from gross or net erosion to dust and the assessment of dust on hot surfaces need to be investigated.• For low-Z materials the build-up of the tritium inventory is dominated by co-deposition with eroded wall atoms.• For W, where erosion and tritium co-deposition are small, the implantation, diffusion and bulk trapping constitute the dominant retention processes. First extrapolations with models based on laboratory data show small contributions to the inventory. For later ITER phases and the extrapolation to DEMO additional tritium trapping sites due to neutron-irradiation damage need to be taken into account.Finally the expected values for erosion and tritium retention are compared to the ITER administrative limits for the lifetime, dust and tritium inventory.

Experiments were carried out in the JET tokamak to determine the critical ion temperature inverse gradient length (R/LTi=R|nablaTi|/Ti) for the onset of ion temperature gradient modes and the stiffness of Ti profiles with respect to deviations from the critical value. Threshold and stiffness have been compared with linear and nonlinear predictions of the gyrokinetic code GS2. Plasmas with higher values of toroidal rotation show a significant increase in R/LTi, which is found to be mainly due to a decrease of the stiffness level. This finding has implications on the extrapolation to future machines of present day results on the role of rotation on confinement.

Abstract:JET underwent a transformation from a full carbon-dominated tokamak to a full metallic device with the ITER-like wall combination for the activated phase with Beryllium main chamber and Tungsten divertor. The ITER-Like Wall (ILW) experiment at JET provides an ideal test bed for ITER and shall demonstrate as primary goals the plasma compatibility with metallic walls and the reduction in fuel retention. We report on a set of experiments ( = 2.0 , = 2.0 − 2.4 , = 0.2 − 0.4) in different confinement and plasma conditions with global gas balance analysis demonstrating a strong reduction of the long term retention rate by a factor ten with respect to carbon references. All experiments have been executed in a series of identical plasma discharges in order to achieve maximum plasma duration until the analysis limit of the active gas handling system has been reached. The composition analysis shows high purity of the recovered gas, typically 99% D. For typical L-mode discharges ( = 0.5 ), type III ( = 5.0 ), and type I ELMy H-mode plasmas ( = 12.0 ) a drop of the retention rate normalised to the operational time in divertor configuration has been measured from 1.27 × 10 has been obtained with the ILW. The observed reduction by one order of magnitude confirms the expected predictions concerning the plasma-facing material change in ITER and widens the operation without active cleaning in the DT phase in comparison to a full carbon device.