The success of the magnetic confinement fusion program relies on the control of the interaction between the hot confined plasma, where the fusion reactions take place, and the wall of the vacuum vessel in which this plasma is maintained. Currently, this interaction is managed by a hardware and magnetic configuration called the divertor, which aims to concentrate the lost plasma fluxes through a dedicated volume (the divertor volume) towards high flux components (surface components of the divertor). The control of dissipative phenomena in this divertor volume is a critical objective that shall allow maintaining high confinement performances in the core (hot plasma) while maintaining fluxes to the components below technological limits. The WEST tokamak, currently operated at CEA Cadarache, has as its main objective the control of this interaction, in close support with the ITER project. The thesis project aims to improve the physical understanding of the control experiments started on WEST, through advanced experimental analysis, to the optimization of a robust and generic control model that can be deployed on WEST to conduct scenarios representative of ITER conditions. The project will also be part of a very active international context on the subject, both in Europe (EUROfusion Activities), in Asia and in the United States, offering a wide spectrum of visibility and possibilities for collaborations and developments. The results will be published in peer-reviewed journals with possibly high impact factors, and may be presented at international conferences.