This paper provides an investigation of the production/transport properties of W impurity in the edge of experimental advanced superconducting tokamak (EAST) H-mode discharges with upper-singlenull configurations by using DIVIMP Monte-Carlo code. The background plasmas are provided by SOLPS5.0 calculations. Firstly, to address the detailed dependence of W impurity behaviors on plasma conditions in the scraped-off layer (SOL)/divertor region of EAST, two pre-edge localized mode (ELM) cases with divertor plasmas in high-recycling and partially-detached regimes, respectively, are considered and compared. It is found that, due to the competitions between the thermal and friction forces (FFs), in the high-recycling case, large quantities of W impurities are transported to the upstream; while in the partially-detached case, most of the W impurities are located near the inner and outer divertors. Furthermore, the W core contamination in type-I ELMy H-mode plasmas of EAST has been estimated by the DIVIMP-SOLPS5.0 simulations, without the consideration of the release of W impurity from the core due to ELMs. During the ELM, simulations exhibit the low-recycling regime of divertor plasma, and high (several hundred eV) target plasma temperatures together with low SOL collisionality ( SOL * n ∼1-3), lead to an order of magnitude increase in the W core contamination rate. However, in the ELM-recovering phase, with the decrease (increase) of divertor plasma temperature (density), significant increase of divertor retention is obtained due to the remarkable increase of FFs on W ions. Besides, intra-ELM phase approximately contributes to more than 50% of the total W core contamination per an ELM cycle. This work represents a step toward a deeper understanding of the W impurity production/transport in EAST H-mode discharges.
Drift effects on the plasma profiles in the scrape-off layer (SOL) of the experimental advanced superconducting tokamak (EAST) have been numerically investigated using the comprehensive 2D edge modeling package, SOLPS-ITER, based on a generic magnetic-equilibrium with lower single null (LSN) configuration. SOL particle diffusivity (DSOL) has been scanned from large (1.0 m2/s) to extremely low (0.02 m2/s) to gradually highlight the role of the drift-based neoclassical mechanisms in the radial particle transport. To address the impact of magnetic-field-direction on the drift-driven transport, plasma profiles, flows and currents in the SOL of EAST plasmas with the toroidal magnetic field (BT) direction favorable and unfavorable for H-mode access, i.e. with the ion B×B drift pointing towards and away from the active X-point, are simulated and analyzed. Results demonstrated that the drift-driven transport, considered as the key process in the formation of SOL plasma profiles, is magnetic-field-direction dependent and thus SOL flows and currents as well as the SOL widths can be obviously affected by the direction of drifts. With BT changed from the favorable direction to the unfavorable one, the flattening of density radial profile as well as the increase of power decay length, in the SOL, can be achieved and can be further enhanced as the weight of turbulent transport (i.e. DSOL) gets reduced, due to the increased contribution of ion parallel viscosity to the radial ion flow. Particularly, with DSOL ≤ 0.05 m2/s in the simulations, the dominant role of drift-based neoclassical mechanisms in the radial particle transport will lead to the formation of the so-called edge density-shelf in plasmas with unfavorable BT. Power scrape-off width in plasmas with unfavorable-BT is much insensitive to the turbulent transport level and can remain to be relatively high even when DSOL has been decreased to be extremely low. Due to the compressing/widening effect of the drift-driven inward/outward radial particle flow, the simulated power scrape-off width exhibits an in-out asymmetry, which is also magnetic-field-direction dependent. This work represents a step towards deeper understanding of physics mechanisms determining SOL widths in EAST.
Developing advanced magnetic divertor configurations to address the coupling of heat and particle exhaust with impurity control is one of the major challenges currently constraining the further development of fusion research, and therefore has become the focus of extensive attention in recent years. In J-TEXT, several new divertor configurations, including the high-field-side single-null poloidal divertor and the island divertor, as well as their associated fundamental edge divertor plasma physics have recently been investigated. The purpose of this paper is to briefly summarize the latest progress and achievements in this relevant research field on J-TEXT in the past few years.
To facilitate the estimation of heat loads on plasma-facing components in fusion devices in various different magnetic geometries, a heat load proxy model was developed based on anisotropic diffusion. In this work, this model is compared to the so-called field line diffusion approach. To facilitate the evaluation of these models, a novel synthetic camera based approach for obtaining heat load distributions from Monte Carlo samples was also developed and implemented. With the assistance of this synthetic camera, heat load predictions for the Wendelstein 7-X divertor were obtained and compared with infrared camera observations. It was found that the anisotropic diffusion based model achieved a closer match to infrared camera observations, while still being suitable in computational effort for large magnetic configuration database scans.
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