Tokamak Simulation Code Research Articles

Hydrogen operation of iter

Abstract Pure hydrogen plasma will be used for the commissioning and initial operation of ITER. Possible operation region and achievable goals of operation in this phase was surveyed by running the 1.5D tokamak simulation code. It was shown that ITER could operate at a modest performance level to be able to research on important issues which would give direct impacts to the later full DT operation phase. By extending the length of the hydrogen operation, the intial construction cost of ITER can be reduced, however, full 100 MW of additional heating capability will be needed to achieve goals of the hydrogen operation phase.

A fast shutdown technique for large tokamaks

A practical method is proposed for the fast shutdown of a large ignited tokamak. The method consists of injecting a rapid series of 30-45 (6 mm) deuterium pellets doped with a small (0.0005%) concentration of krypton impurity, and simultaneously ramping the plasma current and shaping fields down over a period of several seconds using the poloidal field system. Detailed modelling with the Tokamak Simulation Code using a newly developed pellet mass deposition model shows that this method should terminate the discharge in a controlled and stable way without producing significant numbers of runaway electrons. A partial prototyping of this technique was accomplished in TFTR.

Design calculations for fast plasma position control in Korea Superconducting Tokamak Advanced Research

We present the results of fast time scale plasma control simulations for the proposed Korea Superconducting Tokamak Advanced Research (KSTAR) tokamak. Here, the fast plasma control includes stabilization of the vertical instability and rapid radial position control. As a simulation tool, we use the tokamak simulation code (TSC). We evaluate the power supply requirements for both the vertical and radial position control with a similar approach to that which was employed in the analysis of the tokamak physics experiments (TPX). In particular, we investigate the efficacy of control coil locations to minimize the power supply requirements. From this study, it is shown that two separate sets of control coils, one responsible for vertical control and the other responsible for radial control, are required to provide robust control of both vertical and radial position for all the anticipated plasmas in KSTAR experiments.

Numerical simulation of feedback stabilization of axisymmetric modes in tokamaks using driven halo currents

The Tokamak Simulation Code (TSC) has been used to model a new method of feedback stabilization of the axisymmetric instability in tokamaks using driven halo (or scrape-off layer) currents. The method appears to be feasible for a wide range of plasma edge parameters. It may offer advantages over the more conventional method of controlling this instability when applied in a reactor environment.

Poloidal Field Control for the Tokamak Physics Experiment

Control of the poloidal field (PF) in the Tokamak Physics Experiment (TPX) is critical to achieving its mission of advanced tokamak research. Extensive examination of the plasma equilibrium; plasma start-up; plasma position, shape, and current control; and plasma shape reconstruction have been performed as part of the design process. This paper reports the progress in this area. The PF coils have been designed to produce a wide range of plasmas. Plasma start-up can be achieved for multiple conditions. Fast plasma position control coils inside the vacuum vessel are used for short timescale control of the plasma vertical and radial position. Shape and total plasma-current control are provided by the PF coils over a slower timescale. A new algorithm for shape control of a few critical plasma boundary points is described and used in simulations using the Tokamak Simulation Code. Fast magnetostatic reconstruction of the plasma shape is examined to determine the impact of measurement locations and their quality.

-collapse-induced vertical displacement event in high tokamak disruption

Extremely fast vertical displacement events (VDEs) induced by a strong collapse were found in a vertically elongated , high tokamak with a resistive shell through computer simulations using the tokamak simulation code. Although the plasma current quench which has been shown to be the prime cause of VDEs in a relatively low tokamak (Nakamura Y et al 1996 Nucl. Fusion 36 643), was not observed during the VDE evolution, the observed growth rate of VDEs was almost five times faster than the growth rate of the usual positional instability . The essential mechanism of the -collapse-induced VDE was clarified to be the intense enhancement of positional instability due to a large and sudden degradation of the magnetic field decay n-index in addition to the significant destabilization due to a reduction in the stability index . The radial shift of the magnetic axis caused by the collapse induces eddy currents on the resistive shell, and these eddy currents produce a large degradation of the n-index. It is pointed out that the shell geometry characterizes the VDE dynamics, and that the VDE rate depends strongly both on the magnitude of the collapse and the n-index of the equilibria just before the collapse occurs. The JT-60U vacuum vessel is shown to possess the capability of preventing -collapse-induced VDEs.

Acceleration Mechanism of Vertical Displacement Event and its Amelioration in Tokamak Disruptions

Vertical displacement events (VDEs), which are frequently observed in disruptive discharges of elongated tokamaks, are investigated using the Tokamak Simulation Code. We show that disruption events such as a sudden plasma pressure drop (βP collapse) and the subsequent plasma current quench (Ip quench) can accelerate VDEs due to the adverse destabilizing effect of the resistive shell, which has previously been thought to stabilize VDEs. In a tokamak with a surrounding shell which is asymmetric with respect to the geometric mid-plane, the Ip quench also causes an additional VDE acceleration due to the vertical imbalance of the attractive force. While the shell-geometry characterizes the VDE dynamics, the growth rate of VDEs depends strongly on the magnitude of the β3p collapse, the speed of the Iv quench and the n-index of the plasma equilibrium just before the disruption. An amelioration of Iv quench-induced VDEs was experimentally established in the JT-60U tokamak by optimizing the vertical location of the plasma just prior to the disruption. The JT-60U vacuum vessel is shown to be suitable for preventing the βp collapse-induced VDE.

Mechanism of vertical displacement events in JT-60U disruptive discharges

Enhanced vertical displacement events (VDEs), which are frequently observed in JT-60U disruptive discharges, are investigated using the Tokamak Simulation Code (TSC). The rapid plasma current quench can accelerate the vertical displacement, owing to both the up/down asymmetry of the eddy current distribution arising from the asymmetric geometry of the JT-60U vacuum vessel and the degradation of magnetic field decay index n, leading to high growth rates of positional instability. For a slightly elongated configuration (n=-0.9), the asymmetry of attractive forces on the toroidal plasma plays a dominant role in the VDE mechanism. For a more elongated configuration (n=-1.7), the degradation of field decay index n plays an important role on VDEs, in addition to the effect of asymmetric attractive forces. It is shown that the VDE characteristics of a highly elongated configuration with a rapid plasma current quench can be dominated by the field decay index degradation. It is also pointed out that both the softening of current quenches as was experimentally developed in the JT-60U tokamak, and the optimization of the allowable elongation of the plasma cross-section are critical issues in the development of a general control strategy of discharge termination

Simulations of lower hybrid current drive in shaped tokamaks

A time dependent lower hybrid current drive tokamak simulation code has been developed. This code utilizes the BALDUR tokamak simulation code and permits the study of the interaction of lower hybrid current drive with neutral beam heating in shaped cross-section plasmas. Examples of simulations are shown for a high aspect ratio tokamak which illuminate the effect of cross-section shaping on current drive

Vertical displacement events and halo currents

This review examines results from all non-circular tokamaks with a distinct emphasis on investigations in ASDEX-Upgrade. There a major fraction of the experimental time has been dedicated studying vertical displacement events of single null plasmas over a large range of q-values in an attempt to obtain the scaling of both the displacement dynamics and the splitting of forces between those associated with poloidal and toroidal plasma currents as a function of q and Bt. These studies on different tokamaks are accompanied by simulations with-among other codes-the tokamak simulation code TSC, in a version where halo currents flowing in the plasma scrape-off layer (SOL) evolve self-consistently. The technical consequences of VDEs for the machine design, measures taken and first predictions are discussed. Safety setups that have been developed and possible avoidance strategies are briefly described.

TSC plasma halo simulation of a DIII-D vertical displacement episode

A benchmark of the Tokamak Simulation Code (TSC) plasma halo model has been achieved by calibration against a DIII-D vertical displacement episode (VDE) consisting of vertical drift, thermal quench and current quench. With a suitable halo surrounding the main plasma, the TSC predictions are in good agreement with experimental results for the plasma current decay, plasma trajectory, toroidal and poloidal vessel currents, and for the magnetic probe and flux loop values for the entire VDE. Simulations with no plasma halo yield much faster vertical motion and significantly worse agreement with magnetic and flux loop data than do halo simulations

TSC simulation of Ohmic discharges in TFTR

The Tokamak Simulation Code (TSC) has been used to model the time dependence of several Ohmic discharges in the TFTR experiment. The semi-empirical thermal conductivity model and the sawtooth model in TSC have been refined so that good agreement between the simulation and the experiment is obtained in the electron and ion temperature profiles and in the current profiles for the entire duration of the discharges. Neoclassical resistivity gives good agreement with the measured surface voltage and the rate of poloidal flux consumption

A study of the Tokamak de Varennes plasma during fast current ramp-down: Experiment and simulation results with the TSC code

Simulation results obtained with the axisymmetric Princeton Tokamak Simulation Code TSC are compared with experimental results for the time evolution of the Tokamak de Varennes plasma. The code correctly predicts experimental conditions; of special interest are the favourable comparisons of the simulation with experiments during fast current ramp-down (from 250 kA in 40 ms, three times faster than the L/R time).

Pulse Length Assessment of Compact Ignition Tokamak Designs

A time-dependent zero-dimensional code has been developed to assess the pulse length and auxiliary heating requirements of compact ignition tokamak (CIT) designs. By taking a global approach to the calculation, parametric studies can be easily performed. The accuracy of the procedure is tested by comparison with the tokamak simulation code, which uses theory-based thermal diffusivities. A series of runs is carried out at various levels of energy confinement for each of three possible CIT configurations. It is found that for cases of interest ignition or an energy multiplication factor Q {ge} 7 can be attained within the first half of the planned 5-s flattop with 10 to 40 MW of auxiliary heating. These results are supported by analytic calculations.

Modelling of post-disruptive plasma loss in the Princeton Beta Experiment

The free-boundary, axisymmetric tokamak simulation code TSC is used to model the transport time-scale evolution and positional stability of the Princeton Beta Experiment (PBX). A disruptive thermal quench will cause the plasma column to move inwards in major radius. It is shown that the plasma can then lose axisymmetric stability, causing it to displace exponentially off the midplane, terminating the discharge. The accuracy of the code is verified by modelling several controlled experimental shots in PBX.