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The processes in far scrape‐off layer (SOL) and the plasma interaction with the first‐wall (FW) elements may notably affect the tokamak discharge, since they define fuel and impurity recycling, material erosion and redeposition, wall surface heating, etc. For a long time, the far SOL description in most plasma edge transport codes was insufficient or absent at all, and so particle and heat fluxes onto the FW (except divertor plates) were out of consideration. Recently some codes, for example, SOLEDGE and SOLPS‐ITER are upgraded allowing for the extension of the computational grid up to real walls and for corresponding account of the vacuum vessel shape and all in‐vessel elements. The new release of SOLPS‐ITER (the version 3.2.0) required a development of a new code data structure and new approach to numerical approximation of fluid equations compatible with unstructured non‐orthogonal computational grid. Intensive testing of the new code in different conditions is still required. In the present contribution, such a testing is performed for the EAST disconnected double null (DDN) L‐mode discharge. For the first time, the SOLPS‐ITER 3.2.0 modeling results with drifts and currents turned on are presented, and a comparison to former SOLPS‐ITER version (3.0.8) is performed. The far SOL transport and its effects on the discharge performance are studied by comparing the computational results obtained on several meshes which differ by their width in equatorial midplane. A single null (SN) one (with mesh width limited by distance to the secondary separatrix) was examined versus two DDN meshes (one with actual and one with artificially extended targets to make mesh wider) and a true unstructured (the widest) mesh. The notable difference in results obtained on different meshes appears in those places where plasma density does not vanishes at the computational domain boundaries. For the cases on true unstructured (the widest) mesh, the particle and heat fluxes onto central column, limiters, far SOL part of targets, dome umbrella and other EAST far SOL in‐vessel structures are calculated for the first time by SOLPS‐ITER, allowing assessment of the plasma interaction with those surfaces. [ABSTRACT FROM AUTHOR]

The SOLPS‐ITER3.0.8 code with new Grad‐Zhdanov module was applied for predictive modeling of tritium plasma and deuterium‐tritium mixture in tokamak JET. The moderate power scenario of shot JPN #85278 modeled earlier for deuterium plasma was taken as a starting point. Effect of puffing increase, which was applied in experiment, on divertor conditions and inter‐ELM tungsten sputtering for both working gases and on isotope separation in their mixture is analyzed. Both for attached and for semi‐detached outer targets, the tritium plasma shows moderate increase in SOL width compared to deuterium one. The tungsten sputtering at the outer target in the inter‐ELM period for tritium is an order of magnitude bigger than that for deuterium. The D–T separation in the D–T mixture is moderate. [ABSTRACT FROM AUTHOR]

Beryllium transport modeling was carried out in the trace impurity regime by SOLPS-ITER code for tokamak ITER. It is valuable for developing of diagnostics protection from beryllium deposition. TRIM reflection model and Bogdanskii formula sputtering model were applied. The idea is to show possibility of using this code for beryllium transport. This code has key advantage comparing to other codes, which study beryllium transport, because in it the main plasma and impurities are described self-consistently. Therefore, if beryllium affect the background plasma, only this code can take this effect into account. Test calculations were performed with suppressed sputtering yield by 10 times, to reach trace impurity regime. Main features of behavior of beryllium in the tokamak were studied, however under restriction of static wall conditions. Sources, sinks, fluxes and density distribution were determined. In this test calculation plasma state kept unchanged as expected. Distribution of impinging and absorbed fluxes on the outer divertor plate were established in the position of divertor diagnostics. Main ways for the further improvement are suggested.