Simulation of Radial Electric Field and Internal Transport Barrier Formation in Small Size Divertor Tokamak Plasma Edge

The simulation of the radial electric field shear, which is responsible for L-H transition by means B2SOLPS0.5.2D transport code, gives the dependence of this shear on plasma parameters. Also, as result of uni-directional neutral beam heating, internal transport barrier is formed and ion radial heat flux q _ir starts to decrease. Furthermore, the dependence of radial electric field shear on ion temperature gradient ITG has also investigated.     

Internal Transport Barrier in Edge Plasma of Small Size Divertor Tokamak Using Neutral Beam Injection

We model the internal transport barrier “ITB” in edge plasma of small size divertor tokamak with B2SOLPS0.5.2D fluid transport code. The simulation results demonstrated the following: (1) we control the internal transport barrier by altering the edge particle transport through changes the edge toroidal rotation which agree with the result of Burrell et al. (Edge Pedestal control in quiescent H-mode discharges in DIII-D using co-plus counter-neutral beam injection, Nucl Fusion, 49, 085024 (9pp) in 2009). (2) The radial electric field has neoclassical nature near separatrix with discharge by co-injection NBI. (3) The toroidal plasma viscosity has strong influence on the toroidal velocity.     

Simulation of Small Size Divertor Tokamak Plasma Edge in the L-regime with ion Transport Barrier

Simulations of L-regimes of small size divertor tokamak plasma edge have been performed with the B2SOLPS5.0 2D fluid transport code for wide range parameters. A conclusion has been made that, radial electric field in the vicinity and inside separatrix is near to neoclassical electric field value. The poloidal E × B drifts and compensating parallel fluxes in the scrape off layer are large in the L-regime with ITB due to steeper gradients while the qualitative pattern of the flows is similar to that of the L-mode.     

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.     

Simulation of Small Size Divertor Tokamak Plasma Edge Including Self-Consistent Electric Fields

The B2.SOLPES.0.5.2D code (Braams, Contrib Plasma Phys 36:276, 1996; Rozhansky and Tendler, Rev Plasma Phys 19:147, 1996) is applied for modeling SOL (Scrape off Layer) plasma in the small size divertor tokamak. Detailed distributions of the plasma heat flux and other plasma parameters in SOL, especially at the target plate of the divertor are found by modeling. The modeling results show that most of the electron heat flux and small part of ion heat flux arrive at target plate of the divertor, while, a large part of the ion heat flux and part of electron heat flux arrive at the outer wall. Also analysis of the role of poloidal E × B drifts in the redistribution of edge plasma is fulfilled.     

Current in the Edge of Small Size Divertor Tokamak

The scheme of poloidal (parallel) currents calculation is presented and compared to the two-dimensional code results both for the core region and for the Scrape-Off layer (SOL). A method of reduction of the two-dimensional equation for potential to one-dimensional ordinary differential equation is suggested. Also the impact of neutral beam injection (NBI) in plasma edge of small size divertor tokamak (SSDT) is studied.     

Perpendicular Conductivity and Self-Consistent Electric Fields in Small Size Divertor Tokamak Plasma Edge

The radial electric field of small size divertor tokamak in the vicinity of separtrix is simulated by using B2-SOLPS5.0 2D code, in which the most complete system of transport equations (Rozhansky et al., Nucl Fusion, 41:4, 2001) is solved including all the important perpendicular current and E × B drifts. Simulations demonstrated the following results: (a) It is shown that in the vicinity of the separatrix, the radial potential profile is determined by perpendicular currents (b) since, due to the pressure asymmetry, radial diamagnetic current integrated over the closed flux surface is not automatically zero, additional radial currents balance the diamagnetic current and make the average net current zero. (c) On the closed flux surfaces far from separetrix, where the pressure is almost constant, the calculated parallel currents (toroidal current) agree with Pfirsch–Schlueter currents.     

Modelling of Small Size Divertor Tokamak Plasma Edge in the Higher Confinement Regime

A version of the B2SOLPS0.5.2D fluid transport code is the new version of B2SOLPS fluid transport code, which is suited technique to simulate the edge plasma of small size divertor tokamak in the H- regime. The results of simulation provide the following: (1) the radial electric field inside the transport barrier is consistent with the neoclassical nature of the radial electric field. (2) The absolute value of the radial electric field shear at inner side of internal transport barrier is small and consistent with the value of shear before the L–H transition, while the value of shear at barrier is significantly large. (3) As a result of strong radial electric field shear and strong barrier formation the diffusion coefficient reduced by factor ~3 with respect to L-mode while ion heat conductivity reduced by factor ~22 with respect to L-mode inside the barrier. (4) The toroidal (Parallel) flux is directed along co-current direction as L-mode but at inner side of barrier is significantly large in absolute value. (5) The radial profile of toroidal rotation in vicinity of transition layer is determined by the parameter δ (width of the transition layer) depending on the collisionality and anomalous diffusion coefficient.     

Simulation of Toroidal Rotation Effect on Heat Flux Transport in the Edge Plasma of Small Size Divertor Tokamak

The effect of toroidal rotation on heat flux transport in the edge plasma of small size divertor was simulated by B2SOLP0.5.2D transport code. The main results of simulation shows that, the following: (1) the radial heat flux is strongly influenced by toroidal rotation. (2) The amplification of conduction part of radial heat flux imposes nonresilient profile of ion temperature, under which the effect of toroidal rotation on ion temperature profile is strong. (3) The ion distribution and its gradients are lower for counter-injection neutral beam than for co-injection neutral beam. (4) Reversal of toroidal rotation during using neutral beam injection result in reverses of radial electric field and E × B drift velocity. (5) The toroidal rotation strong influence on the ion temperature scale length of the ion temperature gradient (ITG). (6) Switch on and off all drifts leads to higher change in the ion density distribution in edge plasma of small size divertor tokamak when the unbalance neutral beam injection are considered (7) the comparison between radial heat flux at different momentum input shows that, the radial ion heat flux with larger ion temperature scale length in the case of co-injection neutral beam is larger than the ion heat flux with smaller ion temperature scale length in the case of counter-injection neutral beam.     

Simulation of the Radial Electric Field Shear in the Edge Plasma of Small Size Divertor Tokamak

The simulation of the radial electric field shear, which is responsible for L-H transition by means B2SOLPS0.5.2D transport code, gives the dependence of this shear on plasma parameters. Also, as result of uni-directional neutral beam heating, internal transport barrier is formed and ion radial heat flux q _ir starts to decrease. Furthermore, the dependence of radial electric field shear on ion temperature gradient ITG has also investigated.     

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.     

Effect of Centrifugal Force on Poloidal Distribution of Impurities in the Edge of Small Size Divertor Tokamak

Toroidal rotation of tokamak plasma causes an outward major radial displacement of impurities. Here the physics of this phenomenon is analysis and the distribution of impurities on a magnetic surface is determined for the cases of practical interest.     

Turbulent Transport in the Tokamak Edge Plasma and SOL Region

In this work, the turbulent transport in the edge plasma and Scrape-Off Layer (SOL) region of IR-T1 tokamak at the presence of biased limiter has been investigated and analyzed. The time and radial evolution of floating potential, electric field and turbulent transport have been measured by using two arrays of the Langmuir probes in both the radial and poloidal directions. The analyses have been done by the Fast Fourier Transport method and spectral features of them are obtained with the help of the standard Auto-Correlation technique and modified covariance power spectral density estimate. The probability distribution function and actual transfer function magnitude of the radial and poloidal turbulent transport (Γ_r and Γ_p) have been investigated and compared in the edge plasma and SOL region. Also the histogram of turbulent transport has been analyzed and compared in the edge and SOL at presence of positive limiter biasing. The results show that in the edge plasma poloidal turbulent transport (Γ_p) is about of 60 % more than SOL region whereas radial turbulent transport (Γ_r) is about of 40 % less. During the application of positive biasing, it was found that Γ_r in the IR-T1 reduces by about 80 % in the edge plasma and 45–50 % in the SOL. Increase of Γ_p is about of 50 % after applied positive biasing in the edge while it increases 70 % nearly, in the SOL. Consequently, the improvement in confinement can be obtained for positive limiter biasing.     

The Behavior of Radial Electric Field and Toroidal Rotation in the Edge Plasma of Divertor Tokamak

A new method for describing the nature of radial electric field and its relation with toroidal rotation in edge plasma of small size divertor tokamak is proposed in this work. The expression of radial electric field in the edge plasma of small size divertor tokamak can be divided into two parts. The first part E _r ⁽⁰⁾ is related to electrostatic potential of plasma in edge plasma of this tokamak. The second part E _r ⁽¹⁾ is related to contribution of toroidal rotation of radial current in edge plasma of this tokamak. The results of this work provide the following: (1) A new one-dimensional ordinary differential equation for toroidal velocity is obtained. The one-dimensional ordinary differential equation suggest new tool to explaining tokamak experiments involving measurements of plasma rotation and radial electric field. (2) Also the results of this work shows that, the main contribution to the radial electric field inside separatrix (plasma core) gives the term E _r ⁽¹⁾.     

Evaluation of Electron Energy Distribution Function in the IR-T1 Tokamak Edge Plasma

Electron energy distribution function (EEDF) in the IR-T1 Tokamak edge plasma is measured. In this paper, application of the first derivative method for processing the electron part of I–V characteristics is report. This method is based on the kinetic theory in a non-local approach. Results show that in the confined plasma EEDF is not strictly Maxwellian and is Bi-Maxwellian. Also proposed procedure allows evaluating additional plasma parameters such as plasma potential in Tokamak edge plasma.     

Investigation of Edge Plasma Features in the HL-IM Tokamak

1.IntroductionInordertooptimizetokamakandachievehighperformanceplasmasuchashighcentraldensityandtemperature,goodcoreconfinementetc.,thesimultaneouscontroloftheedgeplasmaandcentraloneiscrucial.Therefore,themeasurementandcontroloftheedgeplajsmaparametersareverysignificanttocentralplasma.EdgeparameterssuchaStheedgedensity,temperatureandspacepotentialareusuallymeasuredbyusingLangmuirprobes.Duetotheprobemeasurementwithhightemporalandspa-tialresolution[1],itcanalsobeusedtoinvestigatetheirfluctuatio…     

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.     

Study of Plasma Instabilities and Effects of Velocity Profile in Internal Transport Barrier

The effects of plasma velocity profiles, that are not monotonous but with maximum/minimum, on plasma instability and transport in an internal transport barrier (ITB) are studied. Ion temperature gradient (ITG) instability is investigated. Results from the gyro-kinetic theory are presented. It is found that the mode structures and stability prop     

Edge Plasma Behaviors in the HL-1M Tokamak

Edge plasma parameters such as density, temperature and space potential have widely been measured by using Langmuir probes in HL-1M tokamak. Their distinct features in a few kinds of typical discharges were described. The improvement of     

Experimental Observation of the Edge and Divertor Plasma in HL-2A

Edge plasma characteristics were studied by a fast-scanning 4-probe array and a Much/Reynolds stress/Langmuir 10-probe movable array in the boundary region. These probes could measure the edge plasma temperature, density, poloidal electric field, radial electric field, Reynolds stress, poloidal rotation velocities and their profiles, which could be obtained by changing the radial positions of the probe array shot by shot. The measured results were used to analyse plasma confinement, turbulent fluctuations and correlations. The fixed flush 3-probe arrays were mounted on the 4-divertor neutralization plates at the same toroidal cross-section in the divertor chamber. These probes were used to measure the profiles of the electron temperature, density and float potential in the divertor chamber. Edge plasma behaviours in both limiter configuration and divertor configuration are compared. The decay lengths of the edge temperature and density were measured and is emphasized for plasma behaviours of the supersonic molecular beam injection and lower hybrid current drive. The dependence of the radial gradient of Reynolds stress on the poloidal flow and the radial gradient of the electric field on turbulent loss are discussed.