Equilibrium Reconstruction and Integration of EFIT with Diagnoses in J-TEXT Tokamak

The EFIT program is integrated with the high resolution laser polarimeter interferometer system(POLARIS), the soft X-ray imaging diagnostic system(SXR) and the electron cyclotron emission radiometer(ECE) in the J-TEXT tokamak. Then some internal information about Faraday angle and the position of safety factor q=1 can be obtained as a constraint to EFIT. The modified EFIT code is used to calculate the internal parameters such as flux function, safety factor q, pressure and current density.     

Recent progress on the J-TEXT three-wave polarimeter-interferometer

The J-TEXT three-wave polarimeter-interferometer system (POLARIS), which measures time-space distribution of electron density and current density, has been optimized with both the optical system and the equilibrium reconstruction method. The phase resolution of a Faraday rotation angle has been improved from 0.1 to 0.06 degree in chords from –0.18 to 0.18 m (plasma minor radius), and the sawtooth oscillation behavior has been detected by Faraday rotation angle measurement. By combining the POLARIS measured data and the equilibrium and fitting code (EFIT), an upgraded equilibrium reconstruction method has been developed, which provides a more accurate temporal and spatial distribution of current density and electron density. By means of the optimized POLARIS and improved equilibrium reconstruction, variations of profiles with increasing density have been carried out, under both Ohmic and electron cyclotron resonance heating discharges.     

Plasma equilibrium calculation in J-TEXT tokamak

Plasma equilibrium has been calculated using an analytical method. The plasma profiles of the current density, safety factor, pressure and magnetic surface function are obtained. The analytical solution of the Grad–Shafranov(GS) equation is obtained by the variable separation method and compared with the computed results of the equilibrium fitting code EFIT.     

Application of 3D MHD equilibrium calculation to RMP experiments in the J-TEXT tokamak

The application of resonant magnetic perturbation(RMP) coils could break the initial axisymmetry and change the magnetic topology in tokamak systems. To understand the plasma equilibrium response to the RMP fields, three-dimensional(3 D) non-linear magnetohydrodynamics equilibrium calculations have been carried out using the HINT code for an RMP field-penetration experiment on J-TEXT. The HINT code does not assume perfectly nested flux surfaces, and is able to consider directly the change of magnetic topology due to the RMP field penetrations. Correlations between 3 D equilibrium calculations and experimental observations are presented. The magnetic topologies calculated by HINT were compared with the field topologies obtained from a vacuum approximation method. It turns out that the effects of redistribution of plasma pressure and current due to the formation of magnetic islands at various resonant rational surfaces should be considered self-consistently for understanding the change of magnetic structure. Such changes include changes in the shape and size of magnetic islands, and the distribution of stochastic fields around the magnetic islands and at the plasma boundary, which plays an important role for plasma-wall interactions.     

The Positions and Number of Flux Loops and Magnetic Probes for the HT—7U Superconducting Tokamak

The plasma shape and other paremeters such as βp,li is important for the tokamak deveice where the palsma has a non-circular cross-section of sufficient elongation.The measuered signals of magnetic probes and flux loops are used to reconstruct the plasma shape and the current profile in device operation and plasma shape feed back control system.So the number and positions of magnetic probes and flux loops provides the basis of the plasma reconstruction.This paper instroduce how to use EFIT code (equilibrium fitting code)to determine the number and positions of the magnetic probes and flux loops.The simulation result is given also.     

Real Time Equilibrium Reconstruction Algorithm in EAST Tokamak

The EAST (HT-7U) superconducting tokamak is a national project of China on fusion research, with a capability of long-pulse (∽1000 s) operation. In order to realize a longduration steady-state operation of EAST,some significant capability of real-time control is required. It would be very crucial to obtain the current profile parameters and the plasma shapes in real time by a flexible control system. As those discharge parameters cannot be directly measured,so a current profile consistent with the magnetohydrodynamic equilibrium should be evaluatedfrom external magnetic measurements, based on a linearized iterative least square method, which can meet the requirements of the measurements. The arithmetic that the EFIT (equilibrium fitting code) is used for reference will be given in this paper and the computational efforts are reduced by parametrizing the current profile linearly in terms of a number of physical parameters.In order to introduce this reconstruction algorithm clearly,the main hardware design will be listed also.     

MHD Equilibrium Configuration Reconstructions for HL-2A Tokamak

The EFIT （Equilibrium Fitting） code is modified for the equilibrium configuration reconstruction in HL-2A. Signals from Langmuir probe （LP） at the divertor targ...     

Numerical simulation of plasma response to externally applied resonant magnetic perturbation on the J-TEXT tokamak

Nonlinear magnetohydrodynamic (MHD) simulations of an equilibrium on the J-TEXT tokamak with applied resonant magnetic perturbations (RMPs) are performed with NIMROD (non-ideal MHD with rotation, open discussion). Numerical simulation of plasma response to RMPs has been developed to investigate magnetic topology, plasma density and rotation profile. The results indicate that the pure applied RMPs can stimulate 2/1 mode as well as 3/1 mode by the toroidal mode coupling, and finally change density profile by particle transport. At the same time, plasma rotation plays an important role during the entire evolution process.     

Toroidal Multipolar Expansion for Fast L-Mode Plasma Boundary Reconstruction in EAST

A new method for plasma boundary reconstruction, based on the toroidal multipolar expansion (TME) scheme, is applied successfully in EAST. TME applies a limited number of toroidal multipolar moments based on toroidal coordinates to treat a two-dimensional problem of axisymmetric plasma equilibrium. The plasma boundary reconstructed by TME is consistent with the results by using EFIT. The method is su±ciently reliable and fast for real time shape control.     

Reconstruction of the Density Profile for the EAST Tokamak Based on Polarimeter/Interferometer and Microwave Reflectometer Systems

A plasma density profile reconstruction procedure based on the Park matrix method has been developed for both circular and elongated plasma configuration on the Experimental Advanced Superconducting Tokamak(EAST).This method incorporates the line integrated electron density measured by the HCN interferometer and polarimeter/interferometer(POINT) system,the equilibrium fit(EFIT) based on magnetic measurements and the edge electron density profile provided by the microwave reflectometer.It is shown that when the magnetic flux surfaces are slightly corrected,the fitting error is less than 5% in comparison with the measurement data.     

Internal Magnetic Configuration Measured by ECE Imaging on EAST Tokamak

ECE imaging （electron cyclotron emission imaging） is an important diagnostic which can give 2D imaging of temperature fluctuation in the core of tokamak. A method based on ECE imaging is introduced which can give the information of the position of magnetic axis and the structure of internal magnetic surface for EAST tokamak. The EFIT equilibrium reconstruction is not reliable due to the absence of important core diagnostic at the initial phase for EAST, so the information given by ECE imaging could help to improve the accuracy of EFIT equilibrium reconstruction.     

A Visible Light Imaging System for the Estimation of Plasma Vertical Displacement in J-TEXT

A wide-viewing-angle visible light imaging system (VLIS) was mounted on the Joint Texas Experimental Tokamak (J-TEXT) to monitor the discharge process. It is proposed that by using the film data recorded the plasma vertical displacement can be estimated. In this paper installation and operation of the VLIS are presented in detailed. The estimated result is further compared with that measured by using an array of magnetic pickup coils. Their consistency verifies that the estimation of the plasma vertical displacement in J-TEXT by using the imaging data is promising.     

Influence of the X-point location on edge plasma transport in the J-TEXT tokamak with a high-field-side single-null divertor

High-density experiments in the high-field-side mid-plane single-null divertor configuration have been performed for the first time on J-TEXT.The experiments show an increase in the highest central channel line-averaged density from 2.73 x 1019 m-3 to 6.49 x 1019 m-3,while the X-point moves away from the target by increasing the divertor coil current.The corresponding Greenwald fraction rises from 0.50 to 0.79.For the impurity transport,the density normalized radiation intensity(absolute extreme ultraviolet and soft x-ray)of the central channel density decreased significantly(＞50％)with an increase in the plasma density.To better understand the underlying physics mechanisms,the 3D edge Monte Carlo code coupled with EIRENE(EMC3-EIRENE)has been implemented for the first time on J-TEXT.The simulation results show good agreement with the experimental findings.As the X-point moves away from the target,the divertor power decay length drops and the scrape-off layer impurity screening effect is enhanced.     

FEQ: a new flux coordinates based equilibrium solver including both magnetic axis and separatrix

Accurate tokamak plasma equilibrium solution in flux coordinates is crucial for many stability and transport studies. Different approaches for dealing with singularities in solving the nonlinear Grad–Shafranov (GS) equation in flux coordinates or also known as straight field line coordinates are proposed in this paper. The GS equation is solved by iterating the position of grids directly in flux coordinates, and hence, no additional errors are introduced due to mapping process for a convergent solution. The singularity at magnetic axis in flux coordinates is removed by using a novel coordinate transform technique. Different from other techniques previously developed, no assumption in boundary condition at magnetic axis is used. This is consistent with the fact that there is no physical boundary at the magnetic axis. A flux coordinate system with poloidal coordinate chosen as the geometric poloidal angle is proposed. It conquers the difficulty in no definition of poloidal coordinate in flux coordinates at separatrix because of the singularity at x-point(s) in a divertor configuration. It also simplifies the process for computing poloidal flux coordinate during the iteration for solving the nonlinear GS equation. Non-uniform grids can be applied in both radial and poloidal coordinates, which allows it to increase the spacial resolution near x-point(s) in a divertor configuration. Based on the model proposed in this paper, a new Flux coordinates based EQuilibrium solver (FEQ) in tokamaks is developed. The numerical solutions from this code agree well with both the analytic Solov'ev solution and the numerical one from the EFIT code for a divertor configuration in the EAST tokamak. This code can be applied for simulating different equilibria with prescribed shape, pressure and current profiles, i.e. including both limiter and divertor configurations, positive triangularity and negative triangularity, different β, arbitrary magnetic shear profile etc. It provides a powerful and convenient fixed-boundary inverse equilibrium solver including both magnetic axis and separatrix in the solution for tokamak researches.     

Inspection procedure assessment using modelling capabilities

Before carrying out inspections on technical components, criteria have to be met to qualify the inspection procedure. The inspection qualification can be performed by technical justification or performance demonstration. The qualification of non-destructive testing methods can be checked on full scale mock-ups where real or realistic defects are implemented. To minimize costs it will become mandatory to model the inspection procedure. In the case of ultrasonic testing, the modelling includes the ultrasonic probe, the pulse-defect interaction and the probe geometry. The applied modelling code is the elastodynamic finite integration technique (EFIT) which includes mode conversion effects. The results of the code are either wave fields, A-scans at different probe positions or complete r.f. data fields. Examples demonstrate the efficiency in modelling angle beam probes which transmit both pressure, shear and subsurface longitudinal waves. These waves are scattered by horizontal or surface breaking cracks. The modelled r.f. data fields are used within the synthetic aperture focusing approach to predict the ultrasonic image which would be obtained in performing the experiment. In addition it will be shown that the EFIT wave fronts for anisotropic homogeneous media are explicitly related to the group velocity in these media.     

Real-time mirror steering for improved closed loop neoclassical tearing mode suppression by electron cyclotron current drive in DIII-D

The development and operation of the neoclassical tearing mode (NTM) avoidance and control system for DIII-D, which uses six sets of real-time steerable mirrors in order to move the electron cyclotron current drive (ECCD) deposition location in plasma, is described. The real-time DIII-D NTM control algorithm residing in the Plasma Control System (PCS) automatically detects an NTM by analysis of the Mirnov diagnostics, employs motional Stark effect (MSE) EFIT MHD equilibrium reconstruction to locate the rational q-surface where the NTM island can be found, then calculates the appropriate mirror position for alignment of the ECCD with the island using ray tracing. The control commands from PCS are sent to the electron cyclotron system to switch on and off or modulate the gyrotrons and to the steerable mirror system to move the steerable mirrors to the requested positions. Successful NTM suppression has been achieved in DIII-D using this control system to rapidly align the NTM island and the ECCD deposition location, and to actively maintain the alignment as plasma conditions change.     

The J-TEXT CODAC system design and implementation

Inspired by the ITER COntrol, Data Access and Communication (CODAC) and ITER instrumentation and control system, J-TEXT tokamak has upgraded its control system with J-TEXT CODAC system. The J-TEXT CODAC system is based on Experimental Physics and Industrial Control System (EPICS). The J-TEXT CODAC system covers everything in the J-TEXT control system including both central and plant control systems, similar to the ITER I&C system. J-TEXT CODAC system is built around a single central control system called Central CODAC system. All the control functions including conventional control, interlock, safety and other common services are supervised by CCS. The J-TEXT CODAC system has been implemented and tested on J-TEXT. It not only tests some of the ideas in ITER CODAC in real life, but also explores the feasibility of new approaches that is unique in J-TEXT CODAC system.     

Preliminary Analysis of Advanced Equilibrium Configuration for the Fusion-Driven Subcritical System

The Fusion-Driven Subcritical System (FDS) is a subcritical nuclear energy system drived by fusion neutron source. In this paper, an advanced plasma configuration for FDS system has been proposed, which aims at high beta, high bootstrap current and good confinement. A fixed-boundary equilibrium code has been used to obtain ideal equilibrium configuration. In order to determine the feasibility of FDS operation, a two-dimensional time-dependent free boundary simulation code has been adopted to simulate time-scale evolution of plasma current profile and boundary position. By analyses, the Reversed Shear mode as the most attractive one has been recommended for the FDS equilibrium configuration design.     

Experimental study of sheath potential coefficient in the J-TEXT tokamak

Sheath potential coefficient α is a key parameter, which is used to estimate plasma potential(V_p)for edge plasma physics study. Recently, a series of experiments has been carried out under hydrogen plasmas in the J-TEXT tokamak with swept probe, which is employed for current–voltage(I–V) characteristic measurement. Electron temperature is evaluated from I–V curve by three-parameter fitting method, and the electron energy probability function shows that electron distribution is Maxwellian both outside and inside of last closed flux surface(LCFS). Plasma potential is obtained by crossing point between I–V exponential fitting curve and electron saturation current extrapolating line, which is in good agreement with first derivative probe technique. The α coefficient profile in the vicinity of the LCFS is obtained, which is in the range of 2.1–3, and decreases from outside to inside of LCFS.