Abstract
W7-X completed its plasma operation in hydrogen with island divertor and inertially cooled test divertor unit (TDU) made of graphite. A substantial set of plasma-facing components (PFCs), including in particular marker target elements, were extracted from the W7-X vessel and analysed post-mortem. The analysis provided key information about underlying plasma–surface interactions (PSI) processes, namely erosion, transport, and deposition as well as fuel retention in the graphite components. The net carbon (C) erosion and deposition distribution on the horizontal target (HT) and vertical target (VT) plates were quantified and related to the plasma time in standard divertor configuration with edge transform ι = 5/5, the dominant magnetic configuration of the two operational phases (OP) with TDU. The operation resulted in integrated high net C erosion rate of 2.8 mg s−1 in OP1.2B over 4809 plasma seconds. Boronisations reduced the net erosion on the HT by about a factor 5.4 with respect to OP1.2A owing to the suppression of oxygen (O). In the case of the VT, high peak net C erosion of 11 μm at the strike line was measured during OP1.2B which converts to 2.5 nm s−1 or 1.4 mg s−1 when related to the exposed area of the target plate and the operational time in standard divertor configuration. PSI modelling with ERO2.0 and WallDYN-3D is applied in an interpretative manner and reproduces the net C erosion and deposition pattern at the target plates determined by different post-mortem analysis techniques. This includes also the 13C tracer deposition from the last experiment of OP1.2B with local 13CH4 injection through a magnetic island in one half module. The experimental findings are used to predict the C erosion, transport, and deposition in the next campaigns aiming in long-pulse operation up to 1800 s and utilising the actively cooled carbon-fibre composite (CFC) divertor currently being installed. The CFC divertor has the same geometrical design as the TDU and extrapolation depends mainly on the applied plasma boundary. Extrapolation from campaign averaged information obtained in OP1.2B reveals a net erosion of 7.6 g per 1800 s for a typical W7-X attached divertor plasma in hydrogen.
Highlights
Wendelstein 7-X (W7-X) [1] is, next to the large helical device (LHD) [2], the currently largest operating stellarator in the world with a plasma volume VP of 30 m3 and a large radius R of 5.5 m
The plasma operation with the graphite test divertor unit (TDU) provided vital information about the operational window and plasma–surface interactions (PSI) processes in W7-X operating with island divertor
Boronisation gettered oxygen in boron layers, reduced dramatically the carbon erosion at the Plasma-facing components (PFCs), and permitted, at reduced impurity content in the plasma, the access to the stellarator-favoured high density operation in the core and functional divertor operation with high neutral compression [58]. These optimised first wall conditions are a prerequisite for the foreseen long-pulse operation in W7-X with the new actively cooled divertor [7] aiming at hydrogen plasmas of 1800 s pulse duration
Summary
Wendelstein 7-X (W7-X) [1] is, next to the large helical device (LHD) [2], the currently largest operating stellarator in the world with a plasma volume VP of 30 m3 and a large radius R of 5.5 m. This contribution includes studies at the main zones of interaction visualised in situ by infra-red (IR) thermography [8] as well as in-vessel by optical inspection of the HT and the VT plates. This results in provision of net C erosion rates for typical attached divertor plasma conditions in standard divertor configuration with TDU. A brief summary and conclusions for OP2 are drawn in the last section 6
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