Description
Reducing the heat load on divertor plates in tokamaks is a key challenge for future fusion devices. One possible route is to modify the divertor magnetic geometry. For this reason, Alternative Divertor Configurations (ADCs) are actively studied. Increasing the plasma-wetted area through poloidal magnetic flux expansion is one of the approaches that can lower the peak heat flux to levels below material limits (around 10 MW/m²).
The aim of this work is to study how increased flux expansion affects the heat load on the outer divertor target in the TCV tokamak, which is well suited for research on advanced divertor configurations. Experimental results are compared with simulations performed using the multi-fluid edge plasma code TECXY. This code allows for short computation times due to its semi-analytical model of neutral particles. The TECXY results are also compared with simulations from the widely used SOLPS-ITER code, which treats neutrals using Monte Carlo methods. This comparison helps to better understand the role of flux expansion by using two different modelling approaches. The effect of plasma drifts is also investigated to evaluate their influence on the simulation results. Finally, the limitations of the TECXY code are discussed in order to identify plasma regimes where the simplified neutral model is sufficient and can significantly reduce computational time.