Overview of co-deposition and fuel inventory in castellated divertor structures at JET

Abstract

The main focus of this work is fuel retention in plasma components of the JET water-cooled Mk-I divertors operated with small tiles, first with carbon fibre composite (CFC) and then with castellated beryllium. Until recently these have been the only large-scale structures of this type used in fusion experiments. Three issues regarding fuel retention and material migration are addressed: (i) accumulation in gaps separating tiles and in the grooves of castellation; (ii) comparison of deposition on carbon and beryllium; (iii) in-depth migration of deuterium into the bulk of CFC. The essential results are summarised as follows: (i) co-deposition occurs up to a few cm deep in the gaps between the Mk-I tiles; (ii) fuel inventory in the CFC tile gaps exceeds that on plasma-facing surfaces by up to a factor of 2; (iii) in gaps between the beryllium tiles from the inner divertor corner the fuel content reaches 30% of that on plasma-facing surfaces, whereas in the grooves of castellation in Be the fuel content is less than 3.0% of that found on the top surface; (iv) fuel inventory on the Be tiles is strongly associated with the carbon co-deposition; (v) the D content measured in the bulk (1.5 mm below the surface) on cleaved CFC tiles exceeds 1 · 10 15 cm À2 . Implications of these results for a next-step device are addressed and the transport mechanism into the gaps is briefly discussed. The results presented here suggest that in a machine with non-carbon walls in the main chamber (as foreseen for ITER) the material transport and subsequent fuel inventory in the castellation would be reduced.

Dynamics of erosion and deposition in tokamaks

“…The crucial role of shadowed areas is illustrated by the deuterium-tritium experiment DTE1 in JET in 1997, after which the vast majority of retained tritium was found in the form of hydrogen rich carbon layers on the cooled louvre structures in the pumping duct of the inner divertor leg [4,5]. In the case of metallic plasma-facing materials, the amount of material transported to remote areas can be significantly reduced [6].…”

Material mixing on plasma-facing components: Compound formation

“…The aim of this paper is to provide an account on material mixing leading to compound formation on two different limiters: castellated bulk metal (so-called macro-brush) and W-coated graphite on the plasma-facing surface. The investigation was focused on composition of graphite surfaces located in gaps between tiles of W-coated blocks and inside castellated grooves of the macro-brush, because these areas are considered as potential traps for vast amount of fuel retained in co-deposits formed in areas shadowed from the direct plasma impact [15]. Strong motivation for studying tungsten reactivity under plasma conditions is also related to the ITER-like wall (ILW) Project at the JET tokamak where divertor tiles will be made of bulk metal lamellae (load bearing tile) and coatings on carbon fibre composites [16].…”

Formation of deuterium–carbon inventories in gaps of plasma facing components

“…Growth rates of hydrocarbon inventories at PFC side faces have been studied so far mainly by analysis of samples, which were exposed during an entire experimental campaign [6][7][8] and therefore provide only limited information on the influence of the different physics parameters. New experiments in TEXTOR, DIII-D and ASDEX Upgrade overcome this limitation by exposure of samples in dedicated plasma discharges using manipulator systems.…”