Large Type-I Edge Localized Modes (ELMs) are completely eliminated with small n = 3 resonant magnetic perturbations (RMP) in low average triangularity, " = 0.26, plasmas and in ITER Similar Shaped (ISS) plasmas, " = 0.53, with ITER relevant collisionalities v e " # 0.2. Significant differences in the RMP requirements and in the properties of the ELM suppressed plasmas are found when comparing the two triangularities. In ISS plasmas, the current required to suppress ELMs is approximately 25% higher than in low average triangularity plasmas. It is also found that the width of the resonant q 95 window required for ELM suppression is smaller in ISS plasmas than in low average triangularity plasmas. An analysis of the positions and widths of resonant magnetic islands across the pedestal region, in the absence of resonant field screening or a self-consistent plasma response, indicates that differences in the shape of the q profile may explain the need for higher RMP coil currents during ELM suppression in ISS plasmas. Changes in the pedestal profiles are compared for each plasma shape as well as with changes in the injected neutral beam power and the RMP amplitude. Implications of these results are discussed in terms of requirements for optimal ELM control coil designs and for establishing the physics basis needed in order to scale this approach to future burning plasma devices such as ITER.

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.

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plasma facing components leading to a reduction of their lifetime and, in some cases, requiring their replacement.Although the main interest of studying runaway plasmas is related to their final deposition, most of the runaway electron studies during disruptions have been devoted to the generation of the runaway current during the disruption current quench. However, during the termination phase of the disruption, when the plasma current and the runaway electrons are lost, conversion of the magnetic energy of the runaway plasma into runaway kinetic energy can occur. This can increase substanti ally the energy fluxes deposited by the runaway electrons on the

An overview of the present status of research toward the final design of the ITER disruption mitigation system (DMS) is given. The ITER DMS is based on massive injection of impurities, in order to radiate the plasma stored energy and mitigate the potentially damaging effects of disruptions. The design of this system will be extremely challenging due to many physics and engineering constraints such as limitations on port access and the amount and species of injected impurities. Additionally, many physics questions relevant to the design of the ITER disruption mitigation system remain unsolved such as the mechanisms for mixing and assimilation of injected impurities during the rapid shutdown and the mechanisms for the subsequent formation and dissipation of runaway electron current.