Modeling of divertor power footprint widths on EAST by SOLPS5.0/B2.5-Eirene

The edge plasma code package SOLPS5.0 is employed to simulate the divertor power footprint widths of the experimental advanced superconducting tokamak(EAST)L-mode and ELM-free H-mode plasmas.The divertor power footprint widths,which consist of the scrape-off layer(SOL)widthλ_q and heat spreading 5,are important physical parameters for edge plasmas.In this work,a plasma current scan is implemented in the simulation to obtain the dependence of the divertor power footprint width on the plasma current I_p.Strong inverse scaling of the SOL width with I_p has been achieved for both L-mode and H-mode plasmas in the forms ofλ_(q,L-mode)=4.98×I_p~(-0.68)andλ_(q,H-mode)=1.86×I_p~(-1.08).Similar trends have also been demonstrated in the study of heat spreading with S_(L-mode)=1.95×I_p~(-0.542)and S_(H-mode)=0.756×I_p~(-0.872).In addition,studies on divertor peak heat load and the magnetic flux expansion factor show that both of them are proportional to plasma current.The simulation work here can act as a way to explore the power footprint widths of future tokamak fusion devices such as ITER and the China Fusion Engineering Test Reactor(CFETR).     

Real-time fast ferrite ICRF tuning system on the Alcator C-Mod tokamak

A real-time ion cyclotron range of frequencies (ICRF) antenna matching system has been successfully implemented on Alcator C-Mod. This is a triple-stub tuning system working at 80 MHz, where one stub acts as a pre-matching stub and the other two stubs use fast ferrite tuners (FFTs) to accomplish fast tuning. It utilizes a digital controller for feedback control (200 μs per iteration) using real-time antenna loading measurements as inputs and the coil currents to the FFT as outputs. The system has achieved and maintained matching for a large range of plasma parameters, including L-mode, H-mode, and plasmas with edge localized modes. It has succeeded in delivering up to 1.85 MW net rf power into H-mode plasmas at maximum voltage of 37 kV on the unmatched side of the matching system.     

Results of ICRF Heating Experiments from the EAST 2010 Campaign

Radio frequency(RF) plasma heating in ion cyclotron range of frequencies(ICRF)was successfully performed on the Experimental Advanced Superconducting Tokamak(EAST).This is mainly because lithium wall conditioning was routinely used to reduce both impurity and hydrogen(H) recycling and to improve the ICRF power absorption.Mainly ICRF heating of the H minority regime at 27 MHz has been applied in deuterium plasmas.The ion cyclotron resonance heating(ICRH) is found to depend strongly on plasma preheating.The ICRH efficiency can be much improved in conjunction with the lower hybrid wave(LHW).Effective ion and electron heating was observed with the H minority heating mode.The increase of the stored energy reached30 kJ in L-mode plasma by using the ICRF power of 1.0 MW alone when the H cyclotron resonance layer was at plasma center.     

Experimental study of density pump-out on EAST

In the experimental advanced superconducting tokamak,density pump-out phenomena were observed by using a multi-channel polarimeter-interferometer system under different heating schemes of ion cyclotron resonant heating,electron cyclotron resonance heating,and neutral beam injection.The density pump-out was also induced with application of resonant magnetic perturbation,accompanied with a degradation of particle confinement.For the comparison analysis in all heating schemes,the typical plasma parameters are plasma current 400 k A,toroidal field 2 T,and line average density 2?×?10~(19)m~(-3).The experimental results show that the degree of pump-out is concerned with electron density and heating power.Low density deuterium low confinement(L-mode) plasmas(3.5?×?10~(19)m~(-3)) show strong pump-out effects.The density pump-out correlated with a significant drop of particle confinement.     

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.     

High Field Tokamaks for Burning Plasma Experiments

A direct consequence of the ELMy H-mode regime of tokamaks is that, for a constant value of the energy gain Q, both the plasma linear dimension and the normalized plasma density and beta are decreasing functions of the toroidal magnetic field. In this paper, starting from the conditions foreseen for the latest versions of ITER, we derive the plasma parameters of three tokamak plasmas with a toroidal magnetic field of 8, 10, and 13 T.     

Plasma and antenna coupling characterization in ICRF-wall conditioning experiments

Ion Cyclotron Wall Conditioning (ICWC) discharges, in pulsed-mode operation, were carried out in the divertor tokamaks ASDEX Upgrade (AUG) and JET to simulate the scenario of ITER wall conditioning at half-field (AUG) and full-field (JET). ICWC-plasma and antenna coupling characterization results obtained during the Ion Cyclotron Resonance Frequency (ICRF)-Wall Conditioning experiments performed in helium-hydrogen mixture in AUG and helium-deuterium mixtures in JET are presented here. Safe operational regimes for optimum ICWC in ITER could be explored for different magnetic fields. Satisfactory antenna coupling in the Mode Conversion scenario along with reproducible generation of ICRF plasmas and reliable wall conditioning were achieved by coupling RF power from one or two ICRF antennas at two (AUG, JET) different resonant frequencies. These results are in qualitative agreement with the predictions of 1-D TOMCAT code. Present study of ICWC indicates towards the beneficial effect of application of an additional (along with toroidal magnetic field) stationary vertical (B_V ? B_T) magnetic field on antenna coupling and plasma parameters. The results obtained from JET and AUG tokamaks, presented in this paper, emphasizes the proposed phenomenological schemes for further development of ICWC in superconducting tokamaks.     

Design of a New Type of Stub Tuner in ICRF Experiment

In the Ion Cyclotron Range of Frequency(ICRF) heating experiment,impedance matching is of great practical significance,because wide variations in antenna loading are observed within the discharge,in tokamaks operating in H-mode.A sudden decrease in antenna loading accompanying the L-mode to H-mode transition typically occurs on a timescale of a few millisec onds,as does the increase in loading at the H- to L-mode transition.Therefore,it is necessary to match dynamically in the transmission line between the generator output and the antenna input connections[1].A new type of stub tuner being developed utilizes the difference in radio-frequency wavelengths between gas and liquid due to different relative dielectric constants.The impedance matching can be adjusted in realtime in an attempt to track the variations in the antenna loading.Since there are no mechanically moving parts in the short ends of stub,the change can be more convenient and safe,moreover,it can withstand higher voltage without breakdown.this system device will be applied in the HT-7 superconductor Tokamak ICRF experiment.     

Recent results on the application of electron cyclotron heating to tokamaks

Many experiments using Electron Cyclotron Heating (ECH) of plasmas in tokamaks have been reported over the past several years. At a power level of 4 MW, ECH has achieved electron temperatures as high as 10 keV in the T-10 tokamak, and the H-mode has been attained in divertor discharges in DIII-D and JFT-2M. Regarding global energy confinement in either L-mode or Hmode, ECH appears to be quite similar in efficiency to neutral injection, but in addition to bulk heating it has been useful for many purposes, including study of local electron heat diffusivity through pulse-modulated heating; suppression of sawteeth, Edge Localized Modes, and other MHD activity; suppression of disruptions; preionization and startup; and current drive. In this paper, progress in these areas which has been reported since the IAEA meeting in 1986 will be summarized.Work supported by U.S. Department of Energy Contract No. DEAC03-89ER51114.     

Optimization Study of ICRF Hydrogen Minority Heating in a Deuterium Plasma of EAST

The full wave TORIC code and the Kinetic Fokker-Planck SSFPQL code are combined to perform self-consistent simulations of the ICRF heating in the EAST 2D magnetic configuration.The combined package is applied to the ICRF hydrogen minority heating in a deuterium plasma with the hydrogen concentration up to 10%.The fast wave propagation and absorption properties,power partitions among the plasma species and the RF driven energetic tails have been analyzed.Meanwhile,in order to optimize the ICRF heating,changing the resonance locations has also been considered in EAST plasmas.     

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.     

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.     

Divertor detachment operation in helium plasmas with ITER-like tungsten divertor in EAST

Detachment in helium (He) discharges has been achieved in the EAST superconducting tokamak equipped with an ITER-like tungsten divertor. This paper presents the experimental observations of divertor detachment achieved by increasing the plasma density in He discharges. During density ramp-up, the particle flux shows a clear rollover, while the electron temperature around the outer strike point is decreasing simultaneously. The divertor detachment also exhibits a significant difference from that observed in comparable deuterium (D) discharges. The density threshold of detachment in the He plasma is higher than that in the D plasma for the same heating power, and increases with the heating power. Moreover, detachment assisted with neon (Ne) seeding was also performed in L- and H-mode plasmas, pointing to the direction for reducing the density threshold of detachment in He operation. However, excessive Ne seeding causes confinement degradation during the divertor detachment phase. The precise feedback control of impurity seeding will be performed in EAST to improve the compatibility of core plasma performance with divertor detachment for future high heating power operations.     

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.     

Theory on dynamic matching with adjustable capacitors for the ICRF system of ASDEX Upgrade

The coupling of electromagnetic waves in the Ion Cyclotron Range of Frequencies (ICRF) is an important method to heat magnetically confined plasmas. Changing plasma conditions, which originate from processes like L-mode to H-mode transition or gas puffing, vary the load impedance of the ICRF antennas. To optimize the power transfer from the radio frequency (RF) generators to the antennas and consequently to the plasma, as well as to protect the RF sources against too high reflected power, a system that matches (i.e. transforms) the antenna input impedance to the impedance required by the generator is necessary. At ASDEX Upgrade this matching system consists of two stub tuners for each antenna, which match the antenna impedance for a value preset before the discharge. The length of the stubs cannot be changed fast enough to compensate plasma variations even on the moderate timescale of the confinement time in ASDEX Upgrade. The use of 3 dB-couplers allows operation even with varying load, at the cost of a reduced power to the plasma.When adjustable capacitors are applied in parallel to the stubs, dynamic matching becomes possible on the tens of ms timescale. The paper describes first the calculation of the required capacitance using transmission line theory. In a second model a minimum search algorithm finds, for a given antenna impedance, the length of the stubs needed for matching, now including the initial values of the capacitors. For the chosen pre-match point in the Smith chart, the range of impedances around this point is calculated for which the voltage standing wave ratio (VSWR) can be lowered below a minimum value by readjusting the capacitors within their maximum and minimum values. The matching range is thereby significantly larger than without the application of adjustable capacitors, at least with a frequency of 30 MHz and 36.5 MHz.     

Solenoid-free Plasma Start-up in HIT-II and NSTX using Transient CHI

A new method of non-inductive startup, referred to as transient coaxial helicity injection (CHI), has been successfully developed on the HIT-II experiment. In this method, a plasma current is rapidly produced by discharging a capacitor bank between coaxial electrodes in the presence of toroidal and poloidal magnetic fields. The initial poloidal field configuration is chosen such that the plasma rapidly expands into the chamber. When the injected current is rapidly decreased, magnetic reconnection occurs near the injection electrodes, with the toroidal plasma current forming closed flux surfaces. On HIT-II, CHI-started plasmas outperform discharges initiated by induction alone and consume fewer volt-seconds. The method has now been successfully applied on NSTX for an unambiguous proof-of-principle demonstration of closed-flux current generation without the use of the central solenoid.     

Long Pulse H-Mode Scenarios Sustained by RF Heating on EAST

EAST has demonstrated its capability of long pulse operation using RF heating(LHCD and ICRF)in past experiments.One key issue to realize the long pulse H-mode experiments is to sustain the plasma current for steady state operation.Based on the calculations of the transport code ONETWO and its coupled RF code GENRAY,two scenarios have been proposed to achieve the 10 s H-mode plasma with Ip=400 kA,     

Numerical study of equilibrium solutions for axisymmetric plasmas with toroidal flow obtained using Solovev approach

Solovev approach of finding equilibrium solutions, which was extended to include the vacuum solutions provided by Zheng, Wooton, and Solano, was found extremely useful for the purpose of shaping studies. Its extension to toroidal equilibria with a general plasma flow was examined theoretically in a companion paper by Chu, Hu and Guo. The only meaningful extension was found for plasmas with a pure toroidal rotation and with a constant Mach number. A set of functions {SOLOVEV_ZWSm} was obtained which fixed location of the magnetic axis for equilibria with quasi-constant current density profile, with toroidal flow at constant Mach number and with specific heat capacity 1. The set {Solovev_ZWSm} should have complete shaping capability for plasma shapes with positive curvature at the boundary; but not for plasmas with negative curvature boundary points, i.e. the doublets or bean shaped tokamaks. We report here extensive numerical studies showing the shaping capability of {Solovev_ZWSm} for plasmas with pure toroidal rotations, including the change in topology of the solution when the rotation mach number changes. Included plasma topology are the sphere (spheromaks); and the tokamaks (including the doublets).     

ICRF Experiments and GAMMA 10 Potential Formation on the Tandem Mirror

Target plasmas, on which the formation of the electrostatic potentials and the improvement of the confinement are studied, are produced with ICRF in the GAMMA 10 tandem mirror. The ion temperature of more than 10 keV has been achieved in relatively low density plasmas. When the strong ICRF heating is applied, it is observed that the high frequency and the low frequency fluctuations are excited and suppress the increase of the plasma parameters. Recently, a new high power gyrotron system has been constructed and the ECRH power in plug extends up to 370 kW. The improvement of the confinement due to the formation of the potential in the axial direction and the strong radial electric field shear has been observed.     

Modelling of Edge Plasma on JT-60U Tokamak by Using SOLPS5.0 code

The edge plasma transport code SOLPS5.0 is used for modelling edge plasmas in the experimental shots on JT-60U tokamak and the proles of the radial particle and heat transport coecients D, e and i along the outer midplane have been obtained by tting the code results to the experimental measurement in L-mode shot 39090 and H-mode shots 37851, 37856. The experimental measurement used for tting includes the proles of electron temperature and density along the outer midplane, the pumping speed, the total particle ux from the core boundary to the computational region and the ux density of neutrals near the outer wall. The modelling and tting results show within the pedestal region in H-mode shots 37851 and 37856 the radial particle transport coecient D has larger drop, but, for L-mode shot 39090, the obvious drop of D and e has not been found.