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.     

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 Low–High Transition by Neutral Beam Injection in Edge Plasma of Small Size Divertor Tokamak

This paper reports simulation of L–H transition by fluid transport code B2SOLPS0.5.2D at low ion plasma density on neutral beam injection (NBI) in the edge plasma of small size divertor tokamak. The simulation provides the following results: (1) the transition is possible at plasma density 2 × 10¹⁹ m^?3 with NBI at temperature heating T^heating 3.62 keV. (2) The simulation predicts the generation of large negative radial electric field E _r, which is thought to help L–H transition during NBI, is suggested in the edge plasma of small size divertor tokamak. (3) The toroidal current density in the edge plasma of small size divertor tokamak is plasma density and direction of NBI dependence. (4) Parallel flux transport by anomalous viscosity (turbulent) through separatrix leads to the variation of toroidal current density.     

Formation of Electron Internal Transport Barrier in Edge Plasma of Small Size Divertor Tokamak

The formation of electron internal transport barrier (EITB) during using counter-neutral beam injection (NBI) heating in the edge plasma of small size divertor tokamak can be simulated by using fluid transport code B2SOLPS0.5.2D. The results of simulations give us the following: (1) Plasma heating with counter-neutral beam injection leads to, strong, parabola type electron internal transport barrier (EITB) was formed in the edge plasma of small size divertor tokamak. (2) In case of plasma heating by counter-neutral beam injection, the radial electric field shear (E _r –gradient) was increased, while electron transport coefficients were reduced in conjunction with the formation of electron internal transport barrier (EITB). (3) The plasma heating by counter-neutral beam injection play significantly role in redistribution of parallel (toroidal) velocity in edge plasma of small size divertor tokamak.     

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.     

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.     

Numerical Study on Ripple Loss of Neutral-Beam-Injected Fast Ions in EAST

A systematic comparative study on the behaviors and loss processes of energetic beam ions in the rippled toroidal field on Experimental Advanced Superconducting Tokamak (EAST) is carried out by numerical simulations. The predicted loss fractions of either co-injected or counter-injected neutral beam ions for typical EAST experiments are 9–16 %; while for low current experiments, the ripple loss domain is enlarged with the increase of the safety factor q, resulting in enhanced beam ions loss. In addition, Counter-injected ions give rise to more lost fractions in the relatively high-energy range, and fewer of them distribute into the core plasma region, which suggest that the co-injection scheme is somewhat preferable for plasma heating. Moreover, the total losses of energetic beam ions in both co-injection and counter-injection geometries are seen to be due entirely to the delayed losses owing to a synergistic effect of collisions and ripple. Finally, potential first wall damage appears to be avoidable for long pulse neutral beam injection scenarios.     

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⁾.     

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.     

Observation and characterization of the effect of electron cyclotron waves on toroidal rotation in EAST L-mode discharges

The change in the toroidal rotation of plasma caused by electron cyclotron wave (ECW) injection has been observed in EAST.It is found that the response of the rotation is similar for all possible ECW toroidal injection angles.The core toroidal rotation velocity increases in the co-current direction along with a rise in the plasma temperature and stored energy.The profile of the electron temperature,ion temperature and toroidal rotation velocity gradually become peaked.The change in toroidal rotation in the core increases with the ECW injection power.Different behavior is observed when the ECWs are injected into low hybrid current drive (LHCD) target plasmas,where the electron temperature and rotation profile become peaked,while the ion temperature profile flattens after ECW injection,suggesting different transport characteristics in energy and momentum.     

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.     

MHD Instabilities and Their Effects on Plasma Confinement in Large Helical Device Plasmas with Intense Neutral Beam Injection

MHD stability of the Large Helical Device (LHD) plasmas produced with intense neutral beam injection is experimentally studied. When the steep pressure gradient near the edge is produced through L-H transition or linear density ramp experiment, interchange-like MHD modes whose rational surface is located very close to the last closed flux surface are strongly excited in a certain discharge condition and affect the plasma transport appreciably. In NBI-heated plasmas produced at low toroidal field, various Alfven eigenmodes are often excited. Bursting toroidal Alfven egenmodes excited by the presence of energetic ions induce appreciable amount of energetic ion loss, but also trigger the formation of internal and edge transport barriers.     

Simulations of first-orbit losses of neutral beam injection(NBI) fast ions on EAST

Simulations of first-orbit losses of neutral beam injection(NBI) fast ions in the EAST tokamak have been studied in detail by using the orbit-following code GYCAVA and the NBI code TGCO. Beam ion losses with the wall boundary are smaller than those with the last closed flux surface boundary. In contrast to heat loads on the wall without radio frequency wave(RFW)antennas, heat loads on the wall with RFW antennas are distributed more locally near the RFW antennas. The direction of the toroidal magnetic field dramatically affects the final positions of lost fast ions, which is related to the magnetic drift. The numerical results on heat loads of beam ions corresponding to different toroidal magnetic fields are qualitatively consistent with the experimental results. Beam ion losses increase with the beam energy for the co-current NBIs and the counter-perpendicular NBI. We have studied the behavior of fast ions produced by a small section neutral beam(beamlet) by using the numerical tool NBIT. The distributions of the loss fraction of beamlet fast ions peaked near the edge of the beam section for the counter-current NBIs, and they are related to the injection angle. This indicates that the first-orbit losses can be reduced by changing the shape of beam cross section.     

Analysis of Divertor Heat Flux with Infrared Thermography During Gas Fuelling in the HL-2A Tokamak

In HL-2A tokamaks, the behavior of heat flux deposited on the divertor targets has been studied during deuterium gas fuelling. The heat flux is reduced significantly after supersonic molecular beam injection （SMBI） fuelling during Ohmic and electron cyclotron resonance heating （ECRH） divertor discharges. The SMBI fuelling causes an increase in the plasma density and this change results in the experienced change of the edge properties. Most of this reduction in divertor target heat flux occurs together with a high plasma radiation region located at near the X-point. The largest reduction in heat flux profiles is observed at the outboard divertor separatrix strike point, while the heat flux far from the strike point remains almost unchanged. In particular, with SMBI multi-pulses gas fuelling, a partially detached divertor regime is observed with a highly radiating region at the X-point. With the onset of the partially detached divertor regime, a sudden drop in both heat flux and power flow on the divertor target is observed. The reduction in power load on the divertor targets is roughly equal to the increase in plasma radiation loss.     

Calculation of Neutral Beam Injected Torque and Its Effective Tangency Major Radius for EAST

Toroidal rotation has been recognized to have significant effects on the transport and magnetohydrodynarnic(MHD) stability of tokamak plasmas.Neutral beam injection(NBI) is the most effective rotation generation method on current,tokamak devices.To estimate the effective injected torque of the first neutral beam injection system on EAST,a simplified analytic method was derived.Calculated beam torque values were validated by those obtained from the NUBEAM code simulation.According to the results,for the collisional torque,the effective tangential radius for torque deposition is close to the beam tangency major radius.However,due to the dielectric property of tokamak plasma,the equivalent tangency major radius of the J×B torque is equal to the average major radius of the magnetic flux surface.The results will be useful for the research of toroidal momentum confinement and the experimental analysis of momentum transport related with NBI on EAST.     

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.     

Multi-scale interaction between tearing modes and micro-turbulence in the HL-2A plasmas

The influence of m/n=2/1(m and n are poloidal and toroidal mode numbers) tearing modes on plasma perpendicular flows and micro-fluctuations has been investigated in HL-2 A neutral beam injection heated L-mode plasmas. It is found that the local perpendicular rotation velocity and turbulence energy are modulated by the alternation between the island X-point and O-point of the naturally rotating tearing modes. Cross-correlation analysis indicates that the modulation of density fluctuations by the tearing mode is not only limited to the island region, but also occurs in the edge region near the last closed flux surface. The turbulence exhibits distinct spectral characteristics inside and outside the island region. In addition, it is observed that the particle flux near the strike point is also significantly impacted by the tearing modes. The experimental evidence reveals that there are strong core-edge interactions between the core tearing modes and the edge transport.     

Measurement of beam power and profile for DNB on HT-7 tokamak

In normal experimental operation, a diagnostic neutral beam (DNB) can produce 6 A of extracted beam current in hydrogen at an energy of 49 keV with a pulse length of 100 ms. Hydrogen and deuterium beams can be produced as well. The diagnostic neutral beam has been added to the diagnostic set so that charge-exchange recombination spectroscopy (CXRS) can be used to acquire ion temperature and rotation. The beam power and beam profile distribution of the DNB injection can be obtained with a thermocouple probe measurement system on the HT-7 superconducting tokamak. The thermocouple probe measurement system with 13 thermocouples crossly distributed on the probe plate was used to measure the temperature rise of each coppery target, so the profile distribution of the ion/neutral beam was obtained by calculation. In this paper, the structure of the probe plate on the DNB for HT-7 tokamak and some measurement results are presented.     

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.     

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.