Modelling of the effect of ELMs on fuel retention at the bulk W divertor of JET

Effect of ELMs on fuel retention at the bulk W target of JET ITER-Like Wall was studied with multi-scale calculations. Plasma input parameters were taken from ELMy H-mode plasma experiment. The energetic intra-ELM fuel particles get implanted and create near-surface defects up to depths of few tens of nm, which act as the main fuel trapping sites during ELMs. Clustering of implantation-induced vacancies were found to take place. The incoming flux of inter-ELM plasma particles increases the different filling levels of trapped fuel in defects. The temperature increase of the W target during the pulse increases the fuel detrapping rate. The inter-ELM fuel particle flux refills the partially emptied trapping sites and fills new sites. This leads to a competing effect on the retention and release rates of the implanted particles. At high temperatures the main retention appeared in larger vacancy clusters due to increased clustering rate.Peer reviewe

Modelling of the effect of ELMs on fuel retention at the bulk W divertor of JET

Effect of ELMs on fuel retention at the bulk W target of JET ITER-Like Wall was studied with multi-scale calculations. Plasma input parameters were taken from ELMy H-mode plasma experiment. The energetic intra-ELM fuel particles get implanted and create near-surface defects up to depths of few tens of nm, which act as the main fuel trapping sites during ELMs. Clustering of implantation-induced vacancies were found to take place. The incoming flux of inter-ELM plasma particles increases the different filling levels of trapped fuel in defects. The temperature increase of the W target during the pulse increases the fuel detrapping rate. The inter-ELM fuel particle flux refills the partially emptied trapping sites and fills new sites. This leads to a competing effect on the retention and release rates of the implanted particles. At high temperatures the main retention appeared in larger vacancy clusters due to increased clustering rate.EURATOM 63305

Modelling of the effect of ELMs on fuel retention at the bulk W divertor of JET

Effect of ELMs on fuel retention at the bulk W target of JET ITER-Like Wall was studied with multi-scale calculations. Plasma input parameters were taken from ELMy H-mode plasma experiment. The energetic intra-ELM fuel particles get implanted and create near-surface defects up to depths of few tens of nm, which act as the main fuel trapping sites during ELMs. Clustering of implantation-induced vacancies were found to take place. The incoming flux of inter-ELM plasma particles increases the different filling levels of trapped fuel in defects. The temperature increase of the W target during the pulse increases the fuel detrapping rate. The inter-ELM fuel particle flux refills the partially emptied trapping sites and fills new sites. This leads to a competing effect on the retention and release rates of the implanted particles. At high temperatures the main retention appeared in larger vacancy clusters due to increased clustering rate

Modelling of the effect of ELMs on fuel retention at the bulk W divertor of JET

Effect of ELMs on fuel retention at the bulk W target of JET ITER-Like Wall was studied with multi-scale calculations. Plasma input parameters were taken from ELMy H-mode plasma experiment. The energetic intra-ELM fuel particles get implanted and create near-surface defects up to depths of few tens of nm, which act as the main fuel trapping sites during ELMs. Clustering of implantation-induced vacancies were found to take place. The incoming flux of inter-ELM plasma particles increases the different filling levels of trapped fuel in defects. The temperature increase of the W target during the pulse increases the fuel detrapping rate. The inter-ELM fuel particle flux refills the partially emptied trapping sites and fills new sites. This leads to a competing effect on the retention and release rates of the implanted particles. At high temperatures the main retention appeared in larger vacancy clusters due to increased clustering rate

Overview of plasma-tungsten surfaces interactions on the divertor test sector in WEST during the C3 and C4 campaigns

Studying the ageing of tungsten monoblocks, their erosion and their fuel inventory is the priority of the WEST post-mortem analyses programme. Actively-cooled ITER-like plasma-facing units (PFUs) and special W-coated marker lower divertor tiles were retrieved from the WEST divertor after the C3 and C4 experimental campaigns to perform ex-situ analyses. The erosion/deposition pattern on the divertor was determined. The deposition is found mainly on the inner side which is covered by layered deposits that increase in thickness in the radial direction from a few hundreds of nm to a maximum of >10 mu m. The deposits are mainly composed of W, O, C, B and D coming from transport of W in the vacuum chamber, oxidized layers and boronizations. Traces of Cu, Fe, Mo, Cr, Ag were also detected. A maximum deposition rate of about 1.4 nm/s was estimated while a minimum campaign-averaged net erosion rate of 0.1 nm/s was measured for the erosion markers at the strike line areas. No assessment of the erosion could be done for the W monoblocks due to a lack of diagnostics. However, the W monoblock edges clearly show traces of damage (melting, cracks) when exposed to the parallel heat flux due to relative misalignment of ITER-like PFUs during assembly. Optical hot spots were also evidenced, confirming the numerical simulations, although their impact on the operation and the lifetime of the components was limited.Peer reviewe

Gross and net erosion balance of plasma-facing materials in full-W tokamaks

Gross and net erosion of tungsten (W) and other plasma-facing materials in the divertor region have been investigated in deuterium (D) and helium (He) plasmas during dedicated experiments in L- and H-mode on ASDEX Upgrade and after full-length experimental campaigns on the WEST tokamak. Net erosion was determined via post-exposure analyses of plasma-exposed samples and full-size wall components, and we conclude that the same approach is applicable to gross erosion if marker structures with sub-millimeter dimensions are used to eliminate the contribution of prompt re-deposition. In H-mode plasmas, gross erosion during ELMs may exceed the situation in inter-ELM conditions by 1-2 orders of magnitude while net erosion is typically higher by a factor of 2-3. The largest impact on net erosion is attributed to the electron temperature while the role of the impurity mixtures is weaker, even though both on ASDEX Upgrade and WEST significant amounts of impurities are present in the edge plasmas. Impurities, on the other hand, will lead to the formation of thick co-deposited layers. We have also noted that with increasing surface roughness, net erosion is strongly suppressed and the growth of co-deposited layers is enhanced. In He plasmas, gross erosion is increased compared to D due to the higher mass and charge states of the plasma particles, resulting from larger energies due to sheath acceleration, but strong impurity fluxes can result in apparent net deposition in the divertor. Our results from ASDEX Upgrade and WEST are comparable and indicate typical net-erosion rates of 0.1-0.4 nm s(-1), excluding the immediate vicinity of the strike-point regions.Peer reviewe

Overview of plasma-tungsten surfaces interactions on the divertor test sector in WEST during the C3 and C4 campaigns

Studying the ageing of tungsten monoblocks, their erosion and their fuel inventory is the priority of the WEST post-mortem analyses programme. Actively-cooled ITER-like plasma-facing units (PFUs) and special W-coated marker lower divertor tiles were retrieved from the WEST divertor after the C3 and C4 experimental campaigns to perform ex-situ analyses. The erosion/deposition pattern on the divertor was determined. The deposition is found mainly on the inner side which is covered by layered deposits that increase in thickness in the radial direction from a few hundreds of nm to a maximum of >10 µm. The deposits are mainly composed of W, O, C, B and D coming from transport of W in the vacuum chamber, oxidized layers and boronizations. Traces of Cu, Fe, Mo, Cr, Ag were also detected. A maximum deposition rate of about 1.4 nm/s was estimated while a minimum campaign-averaged net erosion rate of 0.1 nm/s was measured for the erosion markers at the strike line areas. No assessment of the erosion could be done for the W monoblocks due to a lack of diagnostics. However, the W monoblock edges clearly show traces of damage (melting, cracks) when exposed to the parallel heat flux due to relative misalignment of ITER-like PFUs during assembly. Optical hot spots were also evidenced, confirming the numerical simulations, although their impact on the operation and the lifetime of the components was limited