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The W -for tungsten- environment in steady-state tokamak (WEST) project consists in transforming the French tokamak Tore Supra into an X-point divertor one and aims at testing the ITER technology of actively cooled high heat flux tungsten components during ITER-relevant long pulses. In addition to the integration in the vacuum vessel of a lower divertor and an upper one, this metamorphosis has a strong impact on diagnostics initially present in Tore Supra. Indeed, the introduction of a divertor in the upper part of the machine leads to block the lines of sight (LOS) of diagnostics previously installed in the upper ports. Therefore, new diagnostics playing a paramount role for the safe tokamak operation need to be implemented, such as an infrared (IR) thermography diagnostic for the protection of plasma-facing components (PFCs) and heating antennas, and a visible spectroscopy system for the protection of tungsten PFCs. Besides, several existing diagnostics need to be modified to adapt their LOS to the new divertor targets and plasma shape. This paper presents the main issues linked to the integration of diagnostics deriving from the WEST project. First, the changes involved by the upgrade on the set of diagnostics currently implemented in Tore Supra are described. Then, a particular focus is put on the integration of the two new optical systems, namely the IR system and the visible spectroscopy one. These two systems being located within the vacuum vessel, a special care must be paid to their design, to allow them to withstand their harsh environmental conditions. Other new or updated fundamental diagnostics are also briefly presented. Finally, the resulting new implementation of diagnostics envisaged to reach the scientific and technological goals of the WEST project is described. [ABSTRACT FROM PUBLISHER]

Plasma deposition in narrow gaps between tiles of ITER plasma-facing components during ELMs is simulated by means of particle-in-cell technique. The particle and power loads onto and between the divertor tiles are estimated for a multispecies plasma. We simulate an equal mixture of deuterium and tritium (T) for the main plasma with one impurity, the carbon (C). The aim of this paper is to know whether the C ions can enter the narrow gaps between tiles and with which energy. Due to its radioactivity, the T deposition in the gaps is also of high interest and is estimated. The two basic orientations of the gap with respect to the magnetic field lines, poloidal and toroidal, have been studied. The global penetration of the two hydrogen species into the gap is on the order of the gap width (~0.5 mm), and their own contribution on the total heat flux to the gap is given. The main impurity penetrates into the gap in between 50% to 70% of the total plasma-deposition length but with a power deposited representing less than 1% of the total deposited power. [ABSTRACT FROM AUTHOR]