Investigation of discharges with a negative triangularity plasma configuration is carried out on various tokamaks: TCV, D-IIID, and ASDEX-U. Negative triangularity (NT) experiments in the TCV show a reduction in electron heat transport by a factor of two compared with the positive triangularity D-shaped configurations. Recent experiments on DIII-D with a NT configuration showed improved confinement compared to the positive triangularity (PT) D-shaped plasmas over a range of auxiliary heating powers and, in particular, for the case Te Ti. In addition, it was found that the NT-shaped plasma has the capability to achieve significant normalized β. The purpose of this study is a computational confirmation of the possibility of implementing scenarios with NT discharges in the ohmic heating mode in the T-15MD tokamak with its standard poloidal system. Simulation results show that the poloidal system of the T-15MD tokamak is rather flexible for study of scenarios with D-shaped NT. It is proposed to expand the research program in the T-15MD by including in it the study of scenarios with NT.

The results of testing the current scenarios developed and verified by using the DINA code at the KTM tokamak during the autumn–winter campaign of 2021 are presented. The qualitative difference between the developed scenarios is an increase in the plasma current and discharge duration with a constant breakdown stage, as well as the presence of a software stage of reducing the plasma current before the discharge is completed. In addition, the technological limitations of the power supply sources of the coils and the requirement to minimize the current reversal in the control coils in the zone of near-zero currents were taken into account. The main goal of the scenarios obtained is the realization of a current plateau during the discharge with the presence of a stage of software reduction of the plasma current before the end of the discharge. The works carried out in 2021 became the next significant step toward bringing the KTM unit to design parameters.

The paper presents the results of experimental campaigns conducted in 2022-2023 aimed at improving and enhancing plasma discharge parameters at the KTM tokamak. The experiments achieved a plasma current of approximately 500 kA in a limiter configuration, with a discharge duration of up to 2 s. Plasma discharges were obtained in the ohmic heating mode. The results described in the paper represent the next step toward achieving the plasma discharge design parameters for the KTM.

The article presents the results of testing the discharge scenarios developed using the DINA code on the KTM tokamak during the autumn–winter campaign of 2021. The qualitative difference of the developed scenarios is the increase in the plasma current and discharge duration with the unchanged breakdown stage at the bypass voltage of 6 V as well as the presence of a program stage of plasma current reduction before the end of the discharge. In addition, technological limitations in the form of the absence of current reversal in the inductor winding during the discharge and minimization of current reversal in the control windings in the zone of near-zero currents are taken into account. The works conducted in 2021 are the next significant step toward bringing the KTM unit to its design parameters.

This article is devoted to the development of a software and computing system for calculating the evolution of nominal currents and voltages in the coils of the poloidal magnetic system in order to ensure the equilibrium of the plasma of the T-15 tokamak in a given discharge scenario. The obtained evolutions of currents and voltages are necessary for the subsequent construction of the system of magnetic control of the current, shape, and position of the plasma during the discharge scenario. The physical parameters of the discharge and the required evolution of the plasma boundary during the discharge process are set using the graphical interface. The software package is developed on the basis of the DINA plasma physics code within the Matlab-Simulink system. Examples of using the graphical interface as well as the simulation results of the ohmic scenario with a plasma current of 2 MA are presented.

The work is devoted to the analysis of the influence of poloidal current induced in a tokamak vacuum chamber due to change of paramagnetic plasma properties as a result of thermal quench on the magnitude and distribution of electromagnetic forces in a vacuum chamber. The work was performed numerically using the plasma-physics code DINA for conditions of major disruption in the plasma of ITER and T-15MD. A comparison is made of the electromagnetic effect obtained from the toroidal and poloidal currents induced in the vacuum chamber as a result of thermal quench with estimates of the electromagnetic forces due to only toroidal currents. It is shown that the integral radial component of the electromagnetic force in the elements of the vacuum chamber when the poloidal current is taken into account can be much lower than in the case of only the toroidal current induced in the chamber. The paper presents a model for calculating the poloidal current induced in a vacuum chamber first used in the DINA code.

The work presents some simulation results of discharge scenarios at the KTM facility during the autumn campaign 2019. The parameters for providing the plasma breakdown phase of the tokamak are estimated. The experimental data of plasma discharges in the KTM are confirmed by the numerical simulation results. The discharge 3669 is simulated in the mode of given currents in poloidal field coils. The simulation data show that the plasma current exceeds 100 kA. Some recommendations for the achievement of higher plasma discharge parameters are proposed.

Heating loads on the divertor has been the major issue for the tokamak operation. The innovative negative triangularity plasmas are considered to be of the favorable property for the power handling. Investigating the hot-plasma vertical displacement events of the negative triangularity plasma has an important significance for the future fusion reactors. In this paper, the hot-plasma vertical displacement event of the negative triangularity plasmas are investigated by the DINA code, based on the new medium-sized copper conducting tokamak HL-2M. Simulation results show that: (i) The initial vertical instability, of the negative triangularity configuration, develops much more faster than that of the positive triangularity configurations; (ii) During the hot VDE, larger absolute value of the negative triangularity causes speeder development of the initial vertical instability and severer mechanical loads; (iii) During the hot VDE of the negative triangularity plasma, the most non-uniform distribution of the electromagnetic force almost occurs in the outer part of the vacuum vessel. Therefore, an arc design of this part is recommended; (iv) The initial vertical instability, as well as the electromagnetic loads, can be mitigated, by further optimizing the magnetic structure around the divertor region.

Расчёт наводимых напряжений в обмотках полоидального

The work on the development of magnetic plasma control system of modernized T-15 tokamak [1] has been carried out. The aim of this work was in the modeling of plasma current and shape control in quasi-stationary stage of the discharge using plasma-physics code DINA [2] and regulators developed in the process of this work were utilized. Scenario of the discharge was divided into several time segments, each of which had nearly constant plasma parameters. For each of the time segments an own regulator was constructed using identification method [3], as a result of which linear models were obtained for the plasma parameter responses from voltages applied in to the poloidal magnetic coils. These linear models are used to create linear-quadratic regulators which have to be switched in the prescribed scenario moments. Two control circuits carry out the control, one of them stabilises plasma vertical position, and the other one conrols current and shape of the plasma. Using the developed control system regulators the simulation of the quasi-stationary operation of T15 tokamak has been carried out with strongly variable plasma parameters. Possibilities of the T-15 poloidal magnetic system are examined for the case of the X-point position movement of the diverted plasma configuration on a stationary stage of the discharge.

Building T-15 tokamak highlights the problem of clarifying the discharge scenarios foreseen for this installation. An analysis of the transition to the divertor configuration (0.7-1.2 sec.) and the stationary phase was performed in the works [1-3]. It has been shown that in these stages the vertical plasma instability is suppressed by the feedback system, and that the system of magnetic diagnostics restores the plasma boundary to the specified accuracy if the sensors measurement error does not exceed 1-3 percent. Calculations have shown a good degree of compliance with the results obtained using the DINA and TOKSCEN codes that indicates the study reliability of these stages of the discharge. Thus, the most urgent task today is to study the initial stage of the discharge, from the current breakdown before the start of the process of forming the divertor configuration (approximately 0.7 sec.). The study means: 1) the clarification of the time dependence of the current magnitude in the coils of the poloidal magnetic system, 2) the comparison of the results of calculations on the various codes, 3) the clarification of the restoration accuracy of the plasma boundary in the initial stage of the discharge. The aim of the work was to conduct such an analysis. The calculations use numerical codes DINA [4], TOKSCEN [5] and RPB (restoration of the plasma boundary) [6]. The calculations could find a realistic scenario of an initial stage of the discharge. This work was supported by grant from Russian Scientific Foundation (project 14-22-00193).

Study of the hot-plasma vertical displacement event (VDE) in advanced divertor configurations is of significant importance for ITER and for future fusion reactors. The newly designed, medium-sized copper-conductor machine HL-2M has the capability of generating the second X-point for various advanced divertor configurations. In this paper, effects of the second X-point on the hot VDE in HL-2M are numerically investigated by utilizing the non-linear time-dependent DINA code. The simulation results show that the existence of the second X-point at certain special locations appears to have a better stability in the vertical direction, compared to the standard configuration with the same main plasma parameters. Meanwhile, the peak halo current during the current quench tends to increase as the second X-point changes in the horizontal direction. The same quantity decreases as the second X-point changes in the vertical direction away from the dominant X-point. From the view point of minimizing the halo current, the tripod is better than the standard configuration, followed by the snowflake-plus and the exact snowflake (SF) configuration. The SF-minus is the worst scenario. On the other hand, the tripod configuration, as well as the SF minus configurations, results in relatively higher peak electromagnetic force acting on the vacuum vessel, when compared to other aforementioned configurations.

ПОСТРОЕНИЕ ЛИНЕЙНЫХ МОДЕЛЕЙ МЕТОДОМ ИДЕНТИФИКАЦИИ ПЛАЗМЫ ДЛЯ СИНТЕЗА РЕГУЛЯТОРОВ

In order to provide efficient performance of tokamaks with vertically elongated plasma position, control systems for limited and diverted plasma configuration are required. The accuracy, stability, speed of response, and reliability of plasma position control as well as plasma shape and current control depend on the performance of the control system. Therefore, the problem of the development of such systems is an important and actual task in modern tokamaks. In this study, the measured signals from the magnetic loops and Rogowski coils are used to reconstruct the plasma equilibrium, for which linear models in small deviations are constructed. We apply methods of the H∞-optimization theory to the synthesize control system for vertical and horizontal position of plasma capable to working with structural uncertainty of the models of the plant. These systems are applied to the plasma-physical DINA code which is configured for the tokamak Globus-M plasma. The testing of the developed systems applied to the DINA code with Heaviside step functions have revealed the complex dynamics of plasma magnetic configurations. Being close to the bifurcation point in the parameter space of unstable plasma has made it possible to detect an abrupt change in the X-point position from the top to the bottom and vice versa. Development of the methods for reconstruction of plasma magnetic configurations and experience in designing plasma control systems with feedback for tokamaks provided an opportunity to synthesize new digital controllers for plasma vertical and horizontal position stabilization. It also allowed us to test the synthesized digital controllers in the closed loop of the control system with the DINA code as a nonlinear model of plasma.

The negative triangularity plasmas are considered to be of important significance on the favorable capability of the power handling towards the fusion reactors [1]. For better understanding the VDEs characteristic of the negative triangularity plasma, a negative triangularity tokamak model, symmetric with the newly designed medium-sized copper-conductor machine HL-2M [2], is created in this paper. The VDEs, of the double-null negative triangularity configurations in such model, is compared to that of the corresponding positive triangularity configuration in HL-2M, by the DINA code [3,4]. Result analysis shows that, the negative triangularity tokamak plasma keeps some certain VDEs characteristics, e. g. the halo current, as well as the electromagnetic loads on the vacuum vessel, of the hot VDE is larger than that of the major disruption with the cold VDE. Meanwhile, some unique characteristics are observed. Compared to the corresponding positive triangularity plasmas, for the major disruption case of the negative triangularity plasmas, two peaks are generated in the evolution waveform of the halo current. Furthermore, the electromagnetic loads on the vacuum vessel become much higher. During the hot VDE, the negative triangularity plasmas become more unstable in the vertical direction, which will cause more challenge for the feedback control. Whilst, the halo current, as well as the electromagnetic loads on the vacuum vessel, becomes more severe during the current quench (FIG. 1), which is unfavorable for the engineering design. References [1] Medvedev S.Y. et al 2015 Nucl. Fusion 55 063013 [2] Zheng G.Y. et al 2014 Fusion Eng. Des. 89 2621 [3] Khayrutdinov R R and Lukash V E 1993 J. Comput. Phys 109 193 [4] Lukash V et al 2005 Plasma Devices Oper. 13 143 [5] Lukash V.E. and Khayrutdinov R.R. 1996 Plasma Phys. Rep. 22 91 Figure 1. EM loads comparison between the standard and the negative triangularity configuration during the VDE. MF-O14

During the tokamak operation,variation of the stored energy can cause internal perturbations of the plasma.These perturbations may develop into large-scale vertical movement of the whole column for the vertically elongated tokamak,eventually generating the hot vertical displacement event（VDE）.It will cause considerable damage to the machine.In this work,the hot VDE process due to stored energy perturbations is investigated by a mature non-linear time-evolution code DINA.The influence on the vertical instability,the displacement direction and the electromagnetic loads on in-vessel components during the hot VDE are analyzed.It is shown that a larger perturbation leads to faster development of the vertical instability.Meanwhile the variation of the Shafranov shift,due to the energy change,is related to the VDE direction.The vertical electromagnetic force on the vacuum vessel and the halo current flowing in the divertor baffle become larger in the case of VDE moving towards the X point.

Cold and hot vertical displacement events (VDEs) are frequently related to the disruption of vertically-elongated tokamaks. The weak poloidal magnetic field around the null-points of a snowflake divertor configuration may influence the vertical displacement process. In this paper, the major disruption with a cold VDE and the vertical disruption in the HL-2M tokamak are investigated by the DINA code. In order to better illustrate the effect from the weak poloidal field, a double-null snowflake configuration is compared with the standard divertor (SD) configuration under the same plasma parameters. Computational results show that the weak poloidal magnetic field can be partly beneficial for mitigating the vertical instability of the plasma under small perturbations. For major disruption, the peak poloidal halo current fraction is almost the same between the snowflake and the SD configurations. However, this fraction becomes much larger for the snowflake in the event of a hot VDE. Furthermore, during the disruption for a snowflake configuration, the distribution of electromagnetic force on a vacuum vessel gets more non-uniform during the current quench.