Characterisation and optimisation of the divertor of a next step fusion device

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ITER is an experimental device of the tokamak type whose purpose is to demonstrate the scientific and technological feasibility of fusion energy for peaceful purposes. Its construction has been decided and the start of construction in Cadarache, France, by an international partnership consisting of Europe, Japan, Russia, the United States, China, and South Korea is imminent. Canadian groups are expected to participate in specific projects on ITER. Among the most important physics questions for ITER, power and particle control ranks as one of the crucial areas because the divertor, the interface between the plasma and the material plasma-facing components, must simultaneously exhaust the power, the deuterium, and the helium ash at realistic plasma and technological parameters. The key to the definition and optimisation of this vital component and its operating parameters is provided by edge and divertor plasma modelling. Using the state-of-the-art plasma simulation code B2-EIRENE, it is proposed to characterise the plasma parameters of the divertor of a next-step fusion device, ITER, to investigate improved divertor operating scenarios based on the radiation of various injected impurities, to optimise the operating conditions consistent with the other plasma and technological constraints on the device and to investigate the impact of modifications of the divertor geometry. The work will be carried out in a framework of international cooperation.