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AbstractAnalysis of comprehensive investigations of divertor armor erosion under action of edge-localized modes (ELMs) using the PEGASUS-3D code has been performed. Different erosion mechanisms for tungsten (W) and carbon fiber composite (CFC) armor materials have been revealed. This difference explains almost 3 orders of magnitude difference in the erosion rates of CFC and W. Investigation of the crack pattern evolution under repetitive ELM-like surface heat load has been done. Special treatment for tungsten armor tiles has been proposed for reduction of dust production rate in ITER operation regime.

A discussion on impurity production at vertical targets and impurity transport to the core plasma during type I ELMs at the 1 GW m^-2 level is given for ITER-FEAT. It is assumed that ELMs deposit their energy into complex layers with considerable surface roughness and drastically reduced heat conductivity. For impurity production, evaporation from target hot spots as produced during disruptions, and evaporation from levitated dust particles and redeposited material is taken into account. First estimations show that ELM energies have to be below 2 MJ assuming the wetted area as 5 m² in order to arrive at impurity densities close to the x-point well below 10¹⁴ cm^-3 for tungsten and below 1 MJ for impurity densities 10¹⁵ cm^-3 for carbon. Larger impurity densities have the potential to dissipate a considerable fraction of the thermal energy of the core plasma.

A discussion on impurity production at vertical targets and impurity transport to the core plasma during type I ELMs at the 1 GW m^-2 level is given for ITER-FEAT. It is assumed that ELMs deposit their energy into complex layers with considerable surface roughness and drastically reduced heat conductivity. For impurity production, evaporation from target hot spots as produced during disruptions, and evaporation from levitated dust particles and redeposited material is taken into account. First estimations show that ELM energies have to be below 2 MJ assuming the wetted area as 5 m² in order to arrive at impurity densities close to the x-point well below 10¹⁴ cm^-3 for tungsten and below 1 MJ for impurity densities 10¹⁵ cm^-3 for carbon. Larger impurity densities have the potential to dissipate a considerable fraction of the thermal energy of the core plasma.

AbstractThe two-dimensional (2-D) radiation-magnetohydrodynamic (R-MHD) code FOREV-2 was developed for modeling of disruptive hot plasma-wall interaction, for calculation of erosion by evaporation of the ITER-FEAT vertical divertor target, and for calculation of the impurity transport in the divertor. FOREV-2 uses a 2 1/2-D MHD model, a 2-D scheme for anisotropic radiation transport, and a solution of the magnetic field equations in the plasma shield for all three components of the magnetic field. Details of FOREV-2 with emphasis on the MHD equations, the equations for the magnetic field, the vaporization model, the angular dependent multigroup radiation transport, and the optical properties of plasmas, and validation of FOREV-2 against analytical results are discussed.Moreover, disruption simulation experiments, performed at the plasma gun facilities at TRINITI Troitsk were used for validation of FOREV-2 and for investigations of the MHD of typical plasma shields. From the results of the numerical analysis of the simulation experiments, it is concluded that turbulence in the experimental plasma shields is absent and that the stability of the cold and dense part of the plasma shields, which determines the target erosion, can be adequately modeled by FOREV-2 by use of the classical magnetic field diffusion coefficient. The experimentally observed downstream drift of plasma shields along the surface of the vertical targets is due to the lateral motion of a cold and dense plasma layer close to the target. The observed upstream shift of the erosion profiles of vertical targets is due to reradiation from the expanding plasma shield. Line radiation and an appropriate model for anisotropic radiation transport are necessary for a realistic calculation of the reradiation from carbon plasma shields. Moreover, inclusion of the line radiation allows one to get a realistic radiation cooling after switching off the heating. Target erosion in the simulation experiments is caused by radiation. The agreement between calculated and measured erosion for graphite and quartz demonstrates the adequacy of the calculated 2-D radiative target heat loads.The 2-D numerical analysis of the disruption simulation experiments allows one to conclude that such experiments adequately simulate the tokamak plasma shield properties and its dynamics. The extensive validation exercise of FOREV-2 against disruption simulation experiments gives confidence that the numerical analysis of erosion for the ITER-FEAT vertical targets during the thermal quench phase of a disruption and the impurity production during ELMs and its transport toward the x-point to be performed with FOREV-2 is based on sound principles and covers all important aspects of plasma shield behavior and plasma shield stability.

During plasma disruptions and ELMs a plasma shield from vaporized target material is formed in front of the divertor plates within a few microseconds which protects the divertor against further excessive evaporation. For calculations of plasma shield formation and material erosion the one dimensional (1 D) radiation magnetohydrodynamics (MHD) code FOREV-1 with 1 12D MHD model was developed. Experimental results on plasma shield formation obtained at the 2MK-200 disruption simulation facility were used to check the adequacy of the physical models used in FOREV-1. It was demonstrated that the theoretical models adequately describe the complex physical processes during plasma shield formation. Plasma shield formation and material erosion in a tokamak is a 2 D problem requiring a 2 12D MHD model. For modeling of this situation the 2 D code FOREV-2 is under development. First results are presented here.