Simulation of the Radial Electric Field in the Small Size Divertor Tokamak for Discharge with Neutral Beam Injection (NBI)

The radial electric field in the edge plasma of small size divertor tokamak can be simulated by B2SOLPS0.5.2D fluid transport code. The simulation provides the follow results: (1) Switching on and off the part of the parallel plasma viscosity driven by parallel ion diamagnetic heat flux (Bekheit in J. Fusion Energ 27(4), 338–345, 2008; Schneider et al. in Nucl. Fusion 41:387, 2001) and Counter-NBI plasma heating change profile of radial electric field significantly. (2) Switching on and off the parallel plasma viscosity driven by parallel ion diamagnetic heat flux leads to the radial electric field is toroidal magnetic field dependence (3) For the case of counter-NBI plasma heating, the switching on and off the current driven by part parallel plasma viscosity depends on the ion diamagnetic heat flux leads to the ion poloidal velocity is toroidal magnetic field B_T dependence. (4) The profile of the radial electric field in edge plasma of small size divertor tokamak is consistent with poloidal rotation velocity.     

Modeling of Heat Fluxes in Edge Plasma of Small Size Divertor Tokamak

The heat flows out from the tokamak core region are collected on the divertor plates and external wall. Control of heat flux exhaust in the SOL and divertor plates regions is one of the important issues in tokamak physics. There are important phenomena affecting heat flows were simulated. The simulation is based on the B2SOLPS5.0 2D multifluid code. It is demonstrated that, the following results: (1) The simulation shows that, the operation of small size divertor tokamak, the divertor plate with/without impurities influence on profiles of electron, ion temperatures, and heat loads significantly. (2) Under normal direction of parallel (toroidal) magnetic field and different values of edge plasma density, strong “SOL” heat flow exists directed towards the LFS (outer) plate. (3) The simulation results show that, the increasing of the plasma density strong influence on the ion and electron poloidal heat fluxes profile significantly. The ion and electron polodial heat flux increase by factor “~8” and “2.4” times. (4) The simulation results show that the in–out asymmetry of heat fluxes was reversed when switching on/off E × B drifts in the edge plasma of this tokamak. (5) The simulation results show correlation between the in–out asymmetry divertor heat fluxes and E × B drift velocity. (6) The observed heat loads asymmetry between HFS and LFS plates can be explained with the radial electric field in SOL. (7) Also the simulation results performed result in, the in–out asymmetry strong influence on the characteristic length of ion poloidal heat flux.     

ICRF physics aspects of wall conditioning plasma characterization in TEXTOR

This paper focuses on encouraging results obtained on the characterization of RF produced plasmas during pulsed-mode wall conditioning discharges in ion cyclotron resonance frequency (ICRF) regime in the limiter tokamak TEXTOR. Recent Ion Cyclotron Wall Conditioning (ICWC) experiment carried out in TEXTOR tokamak, lead to the identification of various dependences of the antenna-plasma coupling efficiency on the plasma parameters for possible ICWC-discharge cleaning in ITER at half field. Our ICWC experiments emphasize on (i) study of antenna coupling during the mode conversion scenario, (ii) reproducible generation of ICRF plasmas for wall conditioning, by coupling RF power from one or two ICRF antennas and (iii) effect of application of an additional (along with toroidal magnetic field) stationary vertical (B_V ? B_T) or oscillating poloidal magnetic field (B_p ? B_T) on antenna coupling and relevant plasma parameters.     

Theoretical and Experimental Approach in Poloidal Beta and Internal Inductance Measurement on IR-T1 Tokamak

Measurements of poloidal beta β _p and internal inductance l _i are essential in tokamak plasma research. Much more plasma parameters such as the plasma current density profile, magnetohydrodynamics instability, and plasma energy confinement time are determined by using these parameters. Discrete poloidal magnetic probes along with the diamagnetic loop can be utilized in measurement of the plasma poloidal beta β _p and internal inductance l _i . In this paper, theoretical and experimental results in determining β _p and l _i are presented and discussed.     

DIVIMP-B2-EIRENE modelling of 13C migration and deposition in ASDEX Upgrade L-mode plasmas

Carbon transport and migration were studied experimentally and numerically in a high-density, low-confinement mode plasma in the ASDEX Upgrade tokamak. On the last day of plasma operation of the 2004–2005 experimental campaign, ¹³CH₄ was injected into the vacuum vessel from the low field side midplane. A poloidal set of tiles was subsequently removed and analysed for ¹³C deposition. In this work the measured deposition profile is interpreted using the impurity transport code DIVIMP. The simulated poloidal distribution of ¹³C deviates significantly from the measured profile. The simulations indicate that ¹³C is promptly deposited at the wall in the vicinity of the injection port, and is transported to the low field side divertor plate predominately via the scrape-off layer. The B2-EIRENE plasma solution produce stagnant plasma flow in the main scrape-off layer, in contrast to measurements in ASDEX Upgrade and other tokamaks. This is the likely cause of the discrepancy between the measured and the calculated poloidal distribution of the ¹³C deposition. Key model parameters of DIVIMP were varied to determine their effect on the simulated deposition profile.     

Ergodic Magnetic Limiter with Barrier

A barrier is produced to radial diffusion of chaotic field lines and consequently increasing the particles confinement time. The chaotic layer near the plasma edge is created by perturbed Hamiltonian (of order ε) of an ergodic magnetic limiter (EML). Adding a control term of order ε ² to this perturbed Hamiltonian revives invariant tori acting as barriers against plasma particles diffusion. The location of the barrier could be chosen in the chaotic zone of EML at a given place near the edge of plasma column. The chaotic behavior of the magnetic field lines around the barrier is studied by utilizing the maximal Lyapunov exponent, average square displacement (diffusion) and invariant manifolds. The effect of changing the number of EML rings on the barrier is investigated. A barrier is also generated by considering special modes in the Fourier expansion of the perturbed Hamiltonian.     

Progress and improvement of KSTAR plasma control using model-based control simulators

Superconducting tokamaks like KSTAR, EAST and ITER need elaborate magnetic controls mainly due to either the demanding experiment schedule or tighter hardware limitations caused by the superconducting coils. In order to reduce the operation runtime requirements, two types of plasma simulators for the KSTAR plasma control system (PCS) have been developed for improving axisymmetric magnetic controls. The first one is an open-loop type, which can reproduce the control done in an old shot by loading the corresponding diagnostics data and PCS setup. The other one, a closed-loop simulator based on a linear nonrigid plasma model, is designed to simulate dynamic responses of the plasma equilibrium and plasma current (I_p) due to changes of the axisymmetric poloidal field (PF) coil currents, poloidal beta, and internal inductance. The closed-loop simulator is the one that actually can test and enable alteration of the feedback control setup for the next shot. The simulators have been used routinely in 2012 plasma campaign, and the experimental performances of the axisymmetric shape control algorithm are enhanced. Quality of the real-time EFIT has been enhanced by utilizations of the open-loop type. Using the closed-loop type, the decoupling scheme of the plasma current control and axisymmetric shape controls are verified through both the simulations and experiments. By combining with the relay feedback tuning algorithm, the improved controls helped to maintain the shape suitable for longer H-mode (10–16 s) with the number of required commissioning shots largely reduced.     

On the Production of Relativistic Runaway Electrons in Damavand Tokamak

Experimental observations in Damavand tokamak show that hard X-ray is produced by either disruption with I _p  < 20 kA or by shots with I _p  > 20 kA. Hard X-ray also persists from the initiation of plasma discharge to the end. Occurrence of multiple spikes in hard X-ray during the discharge is evident. The propagation of hard X-ray is attributed to runaway electrons. We observe runaway electrons in two regimes with different characteristics. Regime (RADI) is similar to the observations of other Tokamak during disruption on that the plasma current is reduced abruptly and interpreted by Dreicer theory. In the regime of RADII, hard X-ray and subsequently runaway electrons are observed from starting of plasma discharge which provides the condition that the most of runaway electrons contain the toroidal plasma current. Runaway electron beam excites whistler waves and scattered electrons in velocity space and prevent growing the runaway electrons beam.     

Evaluation of applicability of 2D iron core model for two-limb configuration of GOLEM tokamak

This paper presents evaluation of applicability of 2D iron core model for highly non-axisymmetric two limb configuration of GOLEM tokamak (former CASTOR). Presented results explain the long-term discrepancy between measured magnitudes of external poloidal field and those calculated by air-core approach on this tokamak. The model has been applied to two poloidal planes at different toroidal angles in the vacuum vessel region and has shown that close to central column of the transformer, it is possible to correct for 3D effects by variation of chosen dimensions of axisymmetric iron core model. Satisfactory agreement of the 2D model results with the measured distribution of B_R field component was achieved.     

Plasma scenarios, equilibrium configurations and control in the design of FAST

The Fusion Advanced Studies Torus (FAST) conceptual study has been proposed [A. Pizzuto on behalf of the Italian Association, The Fusion Advanced Studies Torus (FAST): a proposal for an ITER Satellite facility in support of the development of fusion energy, in: Proceedings of 22nd IAEA Fusion Energy Conference, Geneva, Switzerland, October 13–18, 2008; Nucl. Fusion, submitted for publication] as possible European ITER Satellite facility with the aim of preparing ITER operation scenarios and helping DEMO design and R&D. Insights into ITER regimes of operation in deuterium plasmas can be obtained from investigations of non linear dynamics that are relevant for the understanding of alpha particle behaviours in burning plasmas by using fast ions accelerated by heating and current drive systems.FAST equilibrium configurations have been designed in order to reproduce those of ITER with scaled plasma current, but still suitable to fulfil plasma conditions for studying burning plasma physics issues in an integrated framework. In this paper we report the plasma scenarios that can be studied on FAST, with emphasis on the aspect of its flexibility in terms of both performance and physics that can be investigated. All plasma equilibria satisfy the following constraints: (a) minimum distance of 3 energy e-folding length (assumed to be 1 cm on the equatorial plane) between plasma and first wall to avoid interaction between plasma and main chamber; (b) maximum current density in the poloidal field coils, transiently, up to around 30 MA/m². The discharge duration is always limited by the heating of the toroidal field coils that are inertially cooled by helium gas at 30 K. The location of the poloidal field coils has been optimized in order to: minimize the magnetic energy; produce enough magnetic flux (up to 35 Wb stored) for the formation and sustainment of each scenario; produce a good field null at the plasma break-down (B_P/B_T < 2 × 10⁻⁴ at low field, i.e. B_T = 4 T and E_T = 2 V/m for at least 40 ms).Plasma position and shape control studies will also be presented. The optimization of the passive shell position slows the vertical stability growth time down to 100 ms.     

Discrete magnetic probes measurement versus analytical method in determining plasma position in IR-T1 Tokamak

One of the analytical solutions to the inhomogeneous Grad–Shafranov equation (GSE) is based on the well-known Solov’ev equilibrium that corresponds to source functions linear in ψ. The GSE has been solved by this method with constraints over plasma current and poloidal beta using rectangular fixed boundary conditions [1]. In this paper a new analytical solution to GSE by imposing constraints on the plasma current and β_p + l_i/2 is presented. This method is used for plasma position determination in IR-T1 Tokamak by considering linear source functions and circular fixed boundary conditions. Plasma position is also measured by discrete magnetic probes and is compared with the analytical technique. Results comparison shows good agreement for a typical discharge in IR-T1 Tokamak.     

Optimization of Damavand Tokamak Poloidal Field Coils Positions and Currents with PSO Algorithm

In order to maintain equilibrium in small or large tokamaks poloidal field coils are utilized, since the function of the poloidal magnetic field is a complex function of current density and the position of the coils, a change in any of the parameters can have a strong effect in the confinement and the magnetohydrodynamic parameters. On the other hand, considering the continuity of the current and the position of the coils, the space being searched is so big that taking all possible conditions into account becomes practically impossible. So a method should be utilized that is able to optimize the position and current of the coils without searching the whole space. This paper seeks to find a new method of deriving the plasma parameter in which a combination of the two methods of neural network and Particles Swarm Optimization is used in order to optimize the position and current of poloidal field coils in Damavand tokamak. Since in the employed methods no special topology is applied, it can be readily used to study any other tokamak.     

Revisiting the poloidal magnetic confinement

This note proposes a closed poloidal magnetic configuration with an in-vessel coil system held by shielded supports. A dipole field is bounded by external coils and constrained into a hollow torus aiming at uniform intensity. In the horizontal mid-plane region the external coils and the dipole outer coils are broken in four arcs and bridged by couple of straight branches. Arcs and straight branches build a set of four side coils. In the clearance between their straight branches four tunnels in the poloidal magnetic field are achieved, to pass the supports and the feeders of the in-vessel coil system.A poloidal machine with a plasma thick as those of present large experiments is outlined. The dipole radius is 5.4 m, the plasma about it has a constant poloidal cross-section about 40 m², a volume about 1300 m³ and a minimum thickness 1 m in the outboard. The magnetic field ranges from 1.4 to 1.8 T.     

Non-inductive Production of ST Plasmas with Washer Gun Sources on the Pegasus Toroidal Experiment

Formation of tokamak-like plasmas via electrostatic helicity injection in the ultra-low aspect ratio Pegasus Toroidal Experiment is reported. Two low-impurity, high-current (1 kA) washer gun current sources have been installed in the lower divertor region. These initially drive current along helical field lines produced by the applied toroidal and vertical fields. At sufficiently low values of externally applied vertical field, the poloidal field generated by the plasma is large enough to cause a poloidal flux reversal. In these cases the plasma relaxes into a tokamak-like configuration. Discharges with I _ϕ≈ 30 kA are produced with less than 2 kA of injected current. These discharges exhibit features indicative of tokamak plasmas, including reversal of poloidal flux at the center column, strong vacuum field deformation, increased current decay times, increased core heating, and characteristic MHD modes common to other helicity-injection-driven toroidal devices.     

Tokamak edge plasma rotation in the presence of the biased electrode

Electrode biasing system was designed, constructed, and installed on the IR-T1 tokamak, and then biasing experiments were carried out. Also, using a Mach probes the effects of radial electric field (produced by biased electrode) on the poloidal and toroidal components of the edge plasma velocity were investigated. The results showed an increase in both toroidal and poloidal components of the edge plasma velocity during biasing regime. Results compared and discussed. During positive biasing, increased E_r tends to slow the poloidal rotation in the electron diamagnetic drift direction, i.e., to speed up rotation in the ion diamagnetic drift direction. An increased toroidal rotation velocity has the opposite effect on the poloidal rotation.     

Radial and Vertical Displacement Events of Plasma in the Damavand Tokamak

The plasma current in tokamak is under the influence of forces in such a way it tents toward the radial expansion. The forces resulting from external self-induced, internal inductance, thermal energy, and magnetic field fluctuations on the plasma column, cause radial expansion. To keep plasma in its position, the Lorentz force should be applied by vertical magnetic field to balance these forces. Control of the plasma position in the radial direction is very complicated. Poloidal beta, β_θ, and the internal inductance parameter depend on plasma current where plasma current parameters themselves are not steady in tokamak. The experimental data of Damavand tokamak is used to compare radial displacement with theoretical prediction. Temporal variation of plasma current along with time variation of R and Z positions of the plasma column is studied. The vertical displacement event takes place because of the elongated cross section of plasma column. Theoretical and experimental results show reasonable agreement.     

MHD Equilibrium and Kink Stability in Damavand Tokamak

Magnetohydrodynamic (MHD) equilibrium is vulnerable to numerous destabilizing mechanisms. Instabilities introduce distortions to the plasma magnetic surfaces and its boundaries, their driving force being the radial gradient of plasma toroidal current density. For certain modal numbers, internal kink modes may develop, and their study is feasible according to the energy principle, in which the change in total potential energy due to the disturbance is evaluated. In this article, we present a totally new analysis of MHD equilibrium and stability, and apply it to Damavand tokamak which has a large aspect ratio. For this purpose, we combine perturbation and Green’s function methods to solve for the equilibrium configuration. At this stage, plasma profiles are found explicitly in terms of Bessel functions, and we present a simple expression for estimation of total toroidal plasma current. Then the rest of plasma profiles, including poloidal magnetic flux, safety factor, and toroidal current density, are obtained and plotted. In the next step, we turn to the stability calculations and show that Damavand plasma is resistant to most of the disturbances.     

B2SOLPS0.5.2D Code Simulations of SOL Flows and Drifts Effects in the Edge Plasma of Small Size Divertor Tokamak

The B2SOLPS0.5.2D code can completely derive measured target asymmetries in edge plasma of small size divertor tokamak (SSDT). SOL flow measurements by the code have been performed in L-mode plasma at various poloidal locations in small size divertor tokamak. The main results of simulations suggest that, the following results: (1) SOLPS0.5.2D simulation predicts J_r^(\textdia) ×B_T J_{r}^{{({\text{dia}})}} \times B_{T} J_r^(\textdia) J_{r}^{{({\text{dia}})}} is diamagnetic current, B _T is normal toroidal magnetic field) force due to the presence of large up-down pressure asymmetries is one of the reasons responsible for observed target asymmetries. (2) The shear of plasma toroidal rotation which is contributed for ITB formation and transition to improved confinement regime is formed near separatrix. The role of centrifugal effect in target asymmetries and SOL flow has been investigated.     

Study on Electron Heat Diffusion Across Stochastic Magnetic Field Affected by Magnitude of Perturbed Magnetic Field in Tokomak

Electron heat diffusion across stochastic magnetic fields is studied numerically in order to find out how the magnitude of perturbed magnetic field affect the enhanced heat conductivity and its radial profile in tokomak plasma physics. For these purposes, non-local stochastic magnetic fields are chosen as research objects in our simulation work. From our numerical results, we can find that the effects of the perturbed magnetic field magnitude are dominated parameter on the enhance electron heat transport conductivity wherever the magnetic field is single island or full stochastic field. Also, a theoretical analysis is provided and compared with numerical results.     

Gravitational flow of a thin film of liquid metal in a strong magnetic field

The influence of a poloidal magnetic field of the spherical Tokamak on super thin (h ≈ 0.1 mm) film flow of liquid metal driven by gravity over the surface of the cooled divertor plate is addressed. The experimental setup developed at the Institute of Physics, University of Latvia (IPUL) is described, which makes it possible to drive and visualize such liquid metal flows in the solenoid of the superconducting magnet “Magdalena”. As applied to the above setup, the magnetic field effect on the operation of the capillary system of liquid metal flow distribution (CSFD) is evaluated by using molten metal (lithium or eutectic InGaSn alloy) with a very small linear flowrate q ≤ 1 mm²/s, spread uniformly across the substrate. The magnetic field effect on the main parameters of the fully developed film flow is estimated for the above-mentioned liquid metals.An approximation technique has been proposed to calculate the development of the gravitational film flow. A non-linear differential second order equation has been derived, which describes the variation of the film flow thickness over the substrate length versus the flowrate q, magnetic field B and the substrate sloping α.Results of InGaSn film flow observations in a strong (B = 4 T) poloidal magnetic field are presented. Analysis of the video records evidences of experimental realization of a stable stationary film flow at width-uniform supply of InGaSn.