This paper presents an analysis of the existing prerequisites and opportunities for the start-up of the Republic of Kazakhstan’s own full-fledged Nuclear Energy Program, a conceptual vision of the development of the Kazakhstani nuclear industry. Recommendations have been formulated for the timely and successful implementation of the program for the introduction of nuclear power plants into the country’s energy generation structure, the development of the necessary related nuclear infrastructure facilities, national legislation, and other issues directly related to nuclear power development.

The pulsed power supply equipment belongs to the basic technological systems intended for implementing the required scenarios of changing currents in magnetic coils. The accuracy of implementing these scenarios directly determines the possibility of attaining the plasma breakdown and the required ultimate plasma parameters. Given the uniqueness of each facility under construction in the world and the installed capacity of the electrotechnical equipment applied in the power supply configuration, one can state with confidence that the construction of similar power supply complexes and their control systems, the optimization of their electrotechnical parameters, and the subsequent accident-free operation are vital tasks in mastering controlled thermonuclear fusion technologies. This paper describes the pulsed power supply system of the KTM tokamak (Kazakhstan) designed for material testing, the digital control system for its power conversion equipment, electrotechnical solutions adopted in the design of the KTM tokamak pulsed power supply system, and findings of tests of some items of equipment and their components. The tests have demonstrated sufficient efficiency of the adopted electrotechnical solutions and the possibility of applying them to implement the pulsed power supply systems for small and medium sized tokamaks.

The Tokamak KTM plant with the aspect ratio of A = 2 and a plasma current of around 750 kA, which is being put into operation in Kurchatov (Kazakhstan), will be able to produce heat loads on the tested samples of materials that are characteristic of the ITER plant, and owing to additional HF heating the plasma pulse duration will reach about 4 s. One of the stages of this work involves studies aimed at obtaining an efficient plasma vertical position control system. The system includes special fast control windings located between the toroidal winding and the inner chamber, and also a fast-acting power supply with feedback. Currently, the tokamak inner chamber with all main systems and windings has been assembled, and a plasma discharge with duration of about 100 ms has been obtained in the limiter configuration. To support operation of the plasma vertical position control system, an own power supply has been designed and assembled, which consists of a thyristor rectifier connected in series with a controlled voltage inverter on IGBT transistors, which has a significant margin in power and response speed. Preliminary computation experiments were carried out in the MATLAB environment, which have confirmed the adequacy of the selected power supply and the system for discharging the energy stored in the windings to the water-cooled ballast resistance. With the remaining tokamak systems having sequentially been put into operation and the first experimental data on plasma motion characteristics in the divertor configuration having become available, it is planned to carry out a series of program experiments aimed at developing and optimizing the feedback system controller algorithm in the power supply of the KTM tokamak fast control windings.

In modern tokamaks, plasma with a vertically elongated cross section is created; however, this configuration is unstable with respect to the vertical position of the plasma, and control systems for the vertical position of the plasma are necessary for the operation of such tokamaks. An important parameter characterizing the capabilities for controlling the vertical position of the plasma is the controllability region. This article assesses the controllability region of the plasma in the KTM tokamak using an HFC coil, taking into account limitations from the power supply. To obtain this assessment, the plasma equilibrium was reconstructed from experimental signals of the tokamak, and a model of plasma motion was constructed on the basis of the reconstructed equilibria. Subsequently, a current control system was simulated for the HFC powered by a voltage inverter in a pulse-width modulation (PWM) mode. The calculated lower estimate of the plasma controllability region was 23 cm, which is sufficiently large for a tokamak with a small plasma radius of 45 cm, and demonstrates the potential for creating an effective plasma position control system using the HFC coil in the KTM tokamak.