Mingzhong Zhao

Understanding helium (He) plasma-induced tungsten (W) surface modifications and the effect of irradiation defects on He plasma-induced W surface modifications under a real divertor environment are important for the operation of fusion reactors. In this study, two iron (Fe) ions pre-irradiated in W samples with dislocation loops and voids, and two unirradiated W samples were exposed to He plasma at the divertor leg position of the Large Helical Device. The gross erosion rate is 1.0 × 10²⁰ atoms m⁻² s⁻¹, 1.0 × 10²⁰ atoms m⁻² s⁻¹, 9.3 × 10¹⁹ atoms m⁻² s⁻¹ and 7.4 × 10¹⁹ atoms m⁻² s⁻¹ for W9, W10, W11 and W12, respectively. The surface of each sample after the exposure was different at the strike point and the two regions on either side of the strike point, the scrape-off layer (SOL) region and the private region. The typical He plasma-induced structures in the SOL region are stripe structure, sawtooth structure and non-undulating structure, which are collectively called He-structures in the present study. At the strike point, the typical He plasma-induced structures are dense W protrusions. In the private flux region, the typical He plasma-induced structures are semi-formed He-structures. The formation of these structures is dependent on the grain orientation. Pinholes were observed in these structures. No significant difference in He plasma-induced structures is found between the pre-irradiated W and the unirradiated W. The formation of non-undulating structure and stripe structure is discussed based on the observed semi-formed He-structures.

In order to clarify the divertor plasma-induced tungsten (W) surface modifications as well as the irradiation defects effect, two kinds of ITER grade W were exposed to the large helical device deuterium (D) plasma at the divertor-leg position. One was the iron (Fe) ion irradiated W to produce irradiation defects, and the other was the unirradiated W. The distributions of divertor plasma-induced surface modifications on these two kinds of W were clarified by scanning electron microscope and transmission electron microscopy. A co-deposition layer which was mainly made up of carbon (C) and Fe was formed at the private flux region (>21 mm). No significant surface change was observed at the strike point region (14–16 mm). The oxygen-enriched amorphous W structures (OEAWs) caused by plasma surface interactions were observed at the 2–14 mm (scrape-off layer region) and 20 mm. At the 2–14 mm and 20 mm, the OEAWs density on the pre-irradiated W sample is lower than that on the unirradiated W sample. On the other hand, the OEAWs size on the pre-irradiated W sample is larger than that on the unirradiated W sample at the 8–12 mm and 20 mm. While, the OEAWs size on the pre-irradiated W sample is smaller than that on the unirradiated W sample at the 2–6 mm. This study implies the possibility of forming OEAWs on the surface of W divertor. And the irradiation defects affect the density and size of OEAWs.

Distributions of deposits and hydrogen (H) on the graphite divertor target elements TM4h4 and TM3v5 in the test divertor units 3 (TDUs3) of Wendelstein 7-X (W7-X) are studied. The TM4h4 and TM3v5 are located at the magnetically symmetric positions in the upper and lower divertor. The microstructure of the deposition layer is characterized by a transmission electron microscope (TEM) combined with a focused ion beam (FIB). Metallic deposits such as iron (Fe), molybdenum (Mo), chromium (Cr) are detected in the deposition layer by energy-dispersive x-ray spectroscopy (EDS). The depth-resolved distribution patterns of boron (B) and metallic deposits on upper and lower horizontal (h) divertor target elements TDUs3-TM4h4 as well as upper and lower vertical (v) divertor target elements TDUs3-TM3v5 are clarified by glow discharge optical emission spectrometry (GDOES). Results for both TDUs3-TM4h4 and TDUs3-TM3v5 show that the B deposition regions exhibit higher H retention due to the co-deposition with deposits. On the other hand, up-down asymmetries in B deposition caused by particle drift exist on both TDUs3-TM4h4 and TDUs3-TM3v5. The B deposition amount on upper TDUs3-TM4h4 is 40% smaller than that on lower TDUs3-TM4h4. While for the vertical target elements, the B deposition amount on upper TDUs3-TM3v5 is 35% larger than that on lower TDUs3-TM3v5. Meanwhile, a shift of around 3 cm in B deposition peaks is observed on upper and lower TDUs3-TM4h4 and TDUs3-TM3v5. Results of numerical simulation of carbon deposition/erosion profiles on the target elements using ERO2.0 code and power flux measured by infrared cameras are shown and compared with the above mentioned B profiles.

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.

The deuterium (D) permeation behavior for 1 displacement per atom Fe²⁺ damaged tungsten (W) was studied by the gas-driven permeation method and compared with undamaged W. The results of thermal desorption spectroscopy showed that dislocation loops and voids were formed in damaged W. It was found that the D permeation behavior in W was affected by irradiation defects. The effective diffusivity and permeability in the damaged W were lower than that in undamaged W. However, the difference in effective diffusivity and permeability between the undamaged sample and the damaged sample was reduced with increasing the heating temperature. Under 965 K, which was enough for D detrapping from voids, the permeability for damaged W was consistent with that for undamaged W.