Feasibility study of divertor heat flux control by lithium injection in KSTAR geometry using SOLPS-ITER modelling

Abstract

For stable and sustainable operation of a magnetic confinement fusion device tokamak, controlling the divertor heat flux is essential. To handle this issue, a method of injecting impurities into the tokamak plasma is used. When lithium is injected as an impurity, the lithium particles tend to stick to the wall when they collide, so it has the advantage of not disturbing the main plasma. In many tokamak devices, studies are being conducted to control the divertor heat flux by using lithium. In this thesis, the physical phenomena that occur when lithium is injected into the KSTAR geometry is analyzed by using SOLPS-ITER modelling, and based on this, the feasibility of divertor heat flux control through lithium injection is evaluated. In addition, by investigating the optimal lithium injection location and rate, the effective divertor heat flux control method is proposed. SOLPS-ITER modelling was performed for mixed plasma of deuterium and lithium, scanning the lithium injection rate in the range of 0-1×〖10〗^22 s^(-1). Four different points within the divertor region were selected as lithium injection location, and various physical phenomena were analyzed, including the changes in divertor heat flux depending on the lithium injection locations. As a result, Li+ and Li2+ ions were locally distributed near the injection location and played the role of major radiators. As the lithium injection rate increased, the radiation power density increased, and the electron temperature decreased in the divertor region of the lithium-injected side (<10 eV near injection location for maximum injection rate 1×〖10〗^22 s^(-2)). Divertor target heat flux density of the lithium-injected side was reduced by up to 40-70% depending on the lithium injection location. By analyzing the poloidal heat flux transported along the SOL flux tube, it was confirmed that heat flux decreased sharply near the lithium injection location. For the cases injecting lithium into the outer divertor, the reduction of the outer divertor heat flux was more effective as the injection location was closer to the SOL flux tube with the maximum outer divertor heat flux density. When the lithium injection location was close to the X-point, lithium particles can enter the inner divertor region, and the inner divertor target heat flux density is reduced for a sufficiently high lithium injection rate. When the lithium injection location was close to the outer divertor target, the density ratio between lithium ions and electrons at the midplane confirmed that lithium particles does not enter the core well. Based on the above results, the optimal lithium injection rate and location for divertor heat flux control were suggested. Also, it is known that the deuterium wall recycling coefficient decreases when lithium is injected into the tokamak plasma, which is an important change that can affect the behavior of divertor plasma. To investigate the effect, the SOLPS-ITER modelling was performed for pure deuterium plasma, scanning the deuterium wall recycling coefficient in the range of 0.84-1.00. As a result, the particle density in the divertor region decreased as the deuterium wall recycling coefficient decreased. In particular, electron and D+ ion densities showed stronger reduction in the inner divertor, while neutral deuterium atom and molecule densities showed stronger reduction in the outer divertor. These in-out asymmetries were explained by the decrease in the deuterium gas puffing rate by density feedback control, and the outboard-distributing tendency of neutral particles due to asymmetry of the KSTAR divertor geometry, respectively. As the deuterium wall recycling coefficient decreased, the radiation power density decreased and the electron temperature increased in the divertor region. The electron temperature showed a stronger increase in the inner divertor region. Peak value of divertor target heat flux density increased from 1.0 MW m^(-2) to 2.0 MW m^(-2) at the inner target, and decreased from 2.5 MW m^(-2) to 1.8 MW m^(-2) at the outer target. Because the particle density and temperature changes of the inner and outer divertor are asymmetric, the heat flux density changes of the inner and outer divertor targets are different. Based on the above studies, using pellet injector or impurity powder dropper (IPD) were proposed as methods of injecting lithium into the actual tokamak divertor plasma. Through modelling and experiments to be performed in future studies, it is expected that the divertor heat flux can be effectively controlled by injecting lithium into the tokamak plasma.