Steady-state Tokamak Plasmas Research Articles

RETRACTED: A new perspective on hydrogen plasma equilibrium calculation along with the MHD potential energy

In this contribution we have presented two techniques for determination of plasma equilibrium position in IR-T1 tokamak: Relaxation and optical techniques. An analysis method of tokamak plasma equilibrium by a relaxation method with specified magnetic axis was presented. The degrees of freedom due to designated positions of the magnetic axis are possible by using poloidal field coil currents. Stable steady state tokamak plasma equilibria are calculated along with the Magnetohydrodynamic (MHD) potential energy. The plasma generates a plasma current which partially or fully cancels the magnetic field from the poloidal field coils. For low-temperature plasmas, the plasma current distribution is centrally peaked; for high-temperature plasmas, the plasma current has a hole. A centrally peaked current distribution in a low-temperature plasma is evolved into a current distribution with a hole by increasing the plasma pressure by Ohmic heating, RF heating, or by neutral beam injection heating. In the second technique, an image processing technique is used for the output signal of CCD camera. Then, plasma emission intensity profile and plasma position were obtained. Results were compared and discussed.

Tokamak plasma equilibrium analysis based on the relaxation method with a specified magnetic axis

In this contribution, we have presented two techniques for the determination of plasma equilibrium position in IR-T1 tokamak: relaxation and optical methods. An analysis method of tokamak plasma equilibrium by a relaxation method with a specified magnetic axis is presented. The degrees of freedom due to designated positions of the magnetic axis are possible by using poloidal field coil currents. Stable steady-state tokamak plasma equilibria are calculated along with the magnetohydrodynamic potential energy. The plasma generates a plasma current which partially or fully cancels the magnetic field from the poloidal field coils. For low-temperature plasmas, the plasma current distribution is centrally peaked; for high-temperature plasmas, the plasma current has a hole. A centrally peaked current distribution in a low-temperature plasma is evolved into a current distribution with a hole by increasing the plasma pressure by Ohmic heating, radio frequency heating, or by neutral beam injection heating. In the second technique, an image-processing technique was used for the output signal of the charge coupled device camera and plasma emission intensity profile and then the plasma position was obtained. Results are compared and discussed.

Investigation of MHD phenomena on Tore Supra by localised eccd perturbation experiments

Electron cyclotron current drive (ECCD) at a low power level has been used on Tore Supra to induce local perturbations of the current density profile. Regimes with strong MHD activity have been analysed, and compared with similar stable discharges, in order to investigate the possible causes of their instability and relate the evolution of the discharge to the localization of EC power deposition. Both co- and counter-current drive pulses have been applied to dominantly or fully non-inductive discharges, sustained by a lower hybrid current drive. Detailed reconstructions by current diffusion calculations have been performed and the error bars evaluated. This method has proved valuable for shedding light on the complex interplay between the evolutions of temperature and safety factor profiles in steady-state tokamak plasmas. The crucial role of the dynamic evolution of rational surfaces has been identified. Moreover, we demonstrate that the operational domain in which ECCD can be employed must cope with the overall current profile characteristics, in particular the position where the safety factor has a minimum.

JET and the Physics Basis of ITER

JET has made unique contributions to the physics basis of ITER by virtue of its ITER-like geometry, large plasma size and D-T capability. The paper discusses recent JET results and their implications for ITER in the areas of standard ELMy H-mode, D-T operation and advanced tokamak modes. In ELMy H-mode the separation of plasma energy into core and pedestal contributions shows that core confinement scales like gyroBohm transport. High triangularity has a beneficial effect on confinement and leads to an integrated plasma performance exceeding the ITER Q = 10 reference case. A revised type I ELM scaling predicts acceptable ELM energy losses for ITER, while progress in physics understanding of NTMs shows how to control them in ITER. The D-T experiments of 1997 have validated ICRF scenarios for heating ITER/a reactor and identified ion minority schemes (e.g. (3He)DT) with strong ion heating. They also show that the slowing down of alpha particles is classical so that the self-heating by fusion alphas should cause no unexpected problems. With the Pellet Enhanced Performance mode of 1988, JET has produced the first advanced tokamak mode, with peaked pressure profiles sustained by reversed magnetic shear and strongly reduced transport. More recently, LHCD has provided easy tuning of reversed shear and reliable access to ITBs. Improved physics understanding shows that rational q-surfaces play a key role in the formation and development of ITBs. The demonstration of real time feedback control of plasma current and pressure profiles opens the path towards fully controlled steady-state tokamak plasmas.

Self-sustained steady state tokamak plasmas

Self-sustained steady-state tokamak equilibria for National Spherical Torus Experiment (NSTX) and Spherical Torus-Volume Neutron Source (ST-VNS) are presented. It is demonstrated that the amount of the pressure gradient driven current which includes the bootstrap current and the diamagnetic current is comparable to the designed current specification in these devices. A possible startup sequence by increasing plasma pressure is also illustrated.

Wall material characteristics in reacting plasma modelling

Using the one-dimensional transport equations for impurities and energy flow in the steady-state tokamak plasma, the self-consistent solutions of the impurity density profile and the temperature profile for a reacting DT plasma with the plasma-surf ace interaction are analyzed. On the basis of these results the influences of the impurities originated from the wall on the central plasma conditions and the Q-value of a reacting plasma are discussed.

Total impurity radiation power losses from steady-state tokamak plasmas

A one-dimensional hydroclynamic-coronal model has been used to calculate the total radiated power loss, Prad from a steady-state tokamak plasma. The results are expressed in terms of quantities routinely measured in most experiments so that scaling with different plasma parameters can readily be seen. Prad is strongly dependent on electron density and temperature in coronal equilibrium (i.e. Prad ∼ with δ ≅ 2.0 and γ between 0.5 and 1.5). The strong anomalous diffusion in many plasmas can, however, considerably reduce these dependences (δ as low as 0.8 and γ between −0.1 and 0.9). Prad can be less than the coronal equilibrium value, but usually it is larger by as much as two orders of magnitude, depending on the diffusion coefficient. A criterion has been developed for the applicability of coronal equilibrium to calculate Prad. Also, expression of the calculations in a simplified form has indicated an accuracy of about a factor of two when applied to more than a hundred experimental plasmas. These results show Prad is often a dominant loss mechanism and can, therefore, lead to considerable modification of existing scaling laws.