In thermonuclear fusion research using magnetic confinement, the tokamak shows the best results today. However, tokamak operation is inherently pulsed. Recently, so-called advanced scenarios are being developed for steady state operation of tokamak experiments by maximising the self-generated current in the plasma at high plasma pressures. The control of the shape of the current density profile in the plasma is key to improve confinement and stability in these advanced scenarios. This thesis focuses on the modelling of the evolution of the current profile in advanced scenarios at the ASDEX Upgrade tokamak and the JET tokamak. This is used to prepare a model for real-time feedback control of the current density profile. These models are verified by simulations and dedicated experiments in ASDEX Upgrade using current drive by neutral beam injection. The majority of the work presented here is based on simulations with a transport code (ASTRA), which includes a model for the ohmic current, a model for bootstrap current, a model for the current driven by external actuators (neutral beam injection). In addition, a model for energy transport (Weiland transport model) is employed. Simulations are performed for advanced scenarios to validate the models used by comparing to experimental observations. The results show that ASTRA simulations describe the evolution of current density profile and temperature profiles appropriately in advanced scenarios. For modelling of a system used for real-time control, a database is required to calculate transfer functions that describe relationship between input signals (neutral beam power from different beam sources) and output signals (total plasma pressure and current density profile). The ASTRA code is used for the simulations to create the database. Model structures suited for systems with many input and output signals are used to calculate a model for current profile control in ASDEX Upgrade and JET. A validation of identified models is carried out using a simulated step response of the neutral beam sources with ASTRA and dedicated experiments with measurements of the current density profile. Both confirm the validity of the models obtained for current density profile control. However, the observations that with neutral beam injection, in some plasma condition, the changes of the current density profile are not in agreement with model calculations are discussed. The approach developed here is applicable to different actuators for current profile or pressure profile control in existing and future experiments.     «

In thermonuclear fusion research using magnetic confinement, the tokamak shows the best results today. However, tokamak operation is inherently pulsed. Recently, so-called advanced scenarios are being developed for steady state operation of tokamak experiments by maximising the self-generated current in the plasma at high plasma pressures. The control of the shape of the current density profile in the plasma is key to improve confinement and stability in these advanced scenarios. This thesis foc...     »

Der Hauptteil der Arbeit basiert auf Simulationen von Thermonuklearen Plasmen in Tokamaks mit dem Transportcode ASTRA. In diesem Code existieren Modelle für den ohmschen Strom, den vom Druckgradienten getriebener Strom und für Strom der durch externen Quellen (Neutralteilcheninjektion) gegeben ist. Die Ergebnisse zeigen, dass die ASTRA Simulationen die Entwicklung des Stromdichteprofils im Experiment geeignet beschreiben. Eine Datenbank ist mit ASTRA berechnet worden, welche die Beziehung zwischen Eingabegrößen (Heizleistung der verschiedenen Neutralteilcheninjektoren) und Ausgabegrößen (Plasmadruck- und Stromdichteprofil) beschreiebt. Diese Datenbank wird zur Berechnung eines Modells zur Stromprofilkontrolle für ASDEX Upgrade und JET benutzt. Die Beobachtung, dass unter bestimmten Bedingungen die Änderungen im Stromdichteprofil durch die Neutralteilchenheizung nicht in Übereinstimmung mit den Modellrechnungen sind, wird diskutiert.     «

Der Hauptteil der Arbeit basiert auf Simulationen von Thermonuklearen Plasmen in Tokamaks mit dem Transportcode ASTRA. In diesem Code existieren Modelle für den ohmschen Strom, den vom Druckgradienten getriebener Strom und für Strom der durch externen Quellen (Neutralteilcheninjektion) gegeben ist. Die Ergebnisse zeigen, dass die ASTRA Simulationen die Entwicklung des Stromdichteprofils im Experiment geeignet beschreiben. Eine Datenbank ist mit ASTRA berechnet worden, welche die Beziehung zwisch...     »