HL-2A装置偏滤器位形研究

1研究方法 等离子体边界是托卡马克平衡运行时等离子体截面的极向磁通函数的等高线，它决定等离子体位形。位形是托卡马克装置实验和工程设计的重要参数，它是由等离子体电流及其分布以及外极向场线圈电流配置共同决定的。     

HT—7托卡马克中等离子体平衡研究

本文解决了二维轴对称近似下带铁芯的托卡马克中等离子体平衡问题，计算了ＨＴ－７托卡马克中的等离子体平衡位形以及极向场系统的非线性电感和垂直场系数。最后应用Ｋｉｒｃｈｈｏｆｆ方程组和平衡垂直场公式得到了一组等离子体，加热场和垂地直场线圈的电流波形的自洽曲线。     

HT－7托卡马克中等离子体平衡研究

本文解决了二维轴对称近似下带铁芯的托卡马克中等离子体平衡问题，计算了ＨＴ－７托卡马克中的等离子体平衡位形以及极向场系统的非线性电感和垂直场系数。最后应用Ｋｉｒｃｈｈｏｆｆ方程组和平衡垂直场公式得到了一组等离子体、加热场和垂直场线圈的电流波形的自洽曲线。     

HL-2A极向场线圈系统的优化设计

通过对原ＡＳＤＥＸ极向场线圈系统进行改造,优化设计出ＨＬ－２Ａ极向场线圈系统,模拟计算了磁场位形演化并估算了伏秒消耗。改造后的极向场线圈系统能够形成８００ｋＡ的等离子体电流,并能产生拉长截面的等离子体偏滤器的位形。分析了改造后的极向场线圈系统的电磁特征,计算了单零,双零及Ｄ形限制器三种等离子体平衡位形。     

EAST位形磁流体平衡的模拟

对托卡马克平衡反演数值计算代码EFIT进行了改进,使之适用于EAST装置的磁流体平衡研究.用改进后的平衡反演代码EFIT模拟了EAST装置稳态阶段的偏滤器平衡位形.提出极向场线圈电流作为线性方程组主约束条件,期望位形作为从约束条件,求解非对称电流条件下自由边界平衡问题的方法.结果表明,该方法能较为容易地得到非对称电流条件产生的平衡位形,在非对称电流条件下自由边界平衡计算是收敛的.     

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为进一步提高HL-2A装置的放电参数和优化等离子体位形，给出了三种可能的主机改造途径：保留真空室，去掉并调整真空室内部分多极场线圈的局部改造方案；保留真空室，重新布局极向场线圈的中等规模改造方案；重新设计真空室和极向场线圈系统的大规模改造方案。对三种改造方案对放电参数和位形的影响和改造工程的可行性进行了分析比较，重新设计真空室和极向场线圈系统的大规模改造方案是最佳选择。     

EAST装置放电预测的初步研究

用托卡马克模拟程序对实验前期的放电进行了模拟,获得了期望的等离子体演化过程及主要参数波形,如极向场线圈电流、等离子体电流、等离子体位置等。通过等离子体控制系统将模拟获取的参数波形用于实验,开展了等离子体R位置控制、完全程序场控制及RZIp控制下的放电模拟对实验的预测研究。模拟结果与实验吻合较好,表明放电模拟具有一定的可预测性,为今后EAST装置开展更深入的物理实验提供了一定的参考。     

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用托卡马克模拟程序对实验前期的放电进行了模拟,获得了期望的等离子体演化过程及主要参数波形,如极向场线圈电流、等离子体电流、等离子体位置等。通过等离子体控制系统将模拟获取的参数波形用于实验,开展了等离子体R位置控制、完全程序场控制及RZIp控制下的放电模拟对实验的预测研究。模拟结果与实验吻合较好,表明放电模拟具有一定的可预测性,为今后EAST装置开展更深入的物理实验提供了一定的参考。     

开放型偏滤器平衡位形的数值计算

本文叙述自由边界环形等离子体MHD平衡崔序SWEQU的数值计算方法,并用该程序计算了开放型偏滤器平衡位形及相应的极向场线圈电流的配置。     

托卡马克等离子体不同平衡位形下的电子回旋波电流驱动

在给定的等离子体总电流和中心电流密度条件下,数值求解平衡方程,求出不同拉长比和三角形变因子的托卡马克等离子体温度、密度、磁场分布,然后通过求解波迹方程和Fokker-Planck方程,分别计算这些位形下的电子回旋波波迹和电流驱动.结果表明：电子回旋波X模从顶部发射时,随着拉长比的增大,波迹会向弱场侧偏移.电子回旋波X模从弱场侧发射时,电子回旋波在等离子体中传播沉积的功率份额随着拉长比的增大而增加,驱动电流位置随着三角形变因子的增大向等离子体中心移动.驱动电流位置随环向和极向发射角的减小向中心移动,对应的电流密度峰值也变大.     

Tokamak plasma equilibrium analysis based on the relaxation method with a specified magnetic axis

In this contribution, we have presented two techniques for the determination of plasma equilibrium position in IR-T1 tokamak: relaxation and optical methods. An analysis method of tokamak plasma equilibrium by a relaxation method with a specified magnetic axis is presented. The degrees of freedom due to designated positions of the magnetic axis are possible by using poloidal field coil currents. Stable steady-state tokamak plasma equilibria are calculated along with the magnetohydrodynamic potential energy. The plasma generates a plasma current which partially or fully cancels the magnetic field from the poloidal field coils. For low-temperature plasmas, the plasma current distribution is centrally peaked; for high-temperature plasmas, the plasma current has a hole. A centrally peaked current distribution in a low-temperature plasma is evolved into a current distribution with a hole by increasing the plasma pressure by Ohmic heating, radio frequency heating, or by neutral beam injection heating. In the second technique, an image-processing technique was used for the output signal of the charge coupled device camera and plasma emission intensity profile and then the plasma position was obtained. Results are compared and discussed.     

Plasma shape optimization for EAST tokamak using orthogonal method

It is necessary to reduce the currents of poloidal field(PF) coils as small as possible, during the static equilibrium design procedure of Experimental Advanced Superconductive Tokamak(EAST). The quasi-snowflake(QSF) divertor configuration is studied in this paper. Starting from a standard QSF plasma equilibrium, a new QSF equilibrium with 300 kA total plasma current is designed. In order to reduce the currents of PF6 and PF14, the influence of plasma shape on PF coil current distribution is analyzed. A fixed boundary equilibrium solver based on a non-rigid plasma model is used to calculate the flux distribution and PF coil current distribution. Then the plasma shape parameters are studied by the orthogonal method. According to the result, the plasma shape is redefined, and the calculated equilibrium shows that the currents of PF6 and PF14 are reduced by 3.592 kA and 2.773 kA, respectively.     

Computational Study of MHD Equilibrium of Noncircular Tokamak Reactor Plasmas

The MHD equilibrium and stability of noncircular tokamak plasmas limited by a separatrix is studied for reactor size systems. A typical example with a plasma current of 15.8 MA and major radius of 8.1 is presented. The required vertical field is generated by a set of discrete external coils and no conducting shell is included. The detailed equilibrium shape is calculated numerically for a vertical elongated plasma with two stagnation points symmetrically located above and below the midplane as would be required for a system with a poloidal divertor. The plasma height to width ratio is 2, the plasma shape factor is 1.6 and poloidal ? is 2.2. The plasma is locally stable. The general stability criteria with respect to quasi-rigid motions (special kink modes) are calculated numerically and found to be satisfied. Size scaling and the engineering constraints are discussed.     

A divertor plasma configuration design method for tokamaks

The efficient and safe operation of large fusion devices strongly relies on the plasma configuration inside the vacuum chamber.It is important to construct the proper plasma equilibrium with a desired plasma configuration.In order to construct the target configuration,a shape constraint module has been developed in the tokamak simulation code(TSC),which controls the poloidal flux and the magnetic field at several defined control points.It is used to construct the double null,lower single null,and quasi-snowflake configurations for the required target shape and calculate the required PF coils current.The flexibility and practicability of this method have been verified by the simulated results.     

The stability margin on EAST tokamak

The experimental advanced superconducting tokamak (EAST) is the first full superconducting tokamak with a D-shaped cross-sectional plasma presently in operation. Its poloidal coils are relatively far from the plasma due to the necessary thermal isolation from the superconducting magnets, which leads to relatively weaker coupling between plasma and poloidal field. This may cause more di?culties in controlling the vertical instability by using the poloidal coils. The measured growth rates of vertical stability are compared with theoretical calculations, based on a rigid plasma model. Poloidal beta and internal inductance are varied to investigate their effects on the stability margin by changing the values of parameters αn and γn(Howl et al 1992 Phys. Fluids B 4 1724), with plasma shape fixed to be a configuration with k = 1.9 and δ = 0.5. A number of ways of studying the stability margin are investigated. Among them, changing the values of parameters κ and li is shown to be the most effective way to increase the stability margin. Finally, a guideline of stability margin Ms(κ,li,A) to a new discharge scenario showing whether plasmas can be stabilized is also presented in this paper.     

The Topolotron: A High Beta Device with Topological Stability

The concept of topological or structural stability is introduced and its importance in the magnetic confinement of plasmas is discussed. Topological stability requires the presence of a pair of limit cycles in the magnetic field configuration. This paper deals with the design of an experimental device possessing limit cycles. The design includes a high beta (? ? 1), high density (~1016), hot (~100 eV) hygrogen plasma which is to be compressed by a factor of about 5 in a toroidal device of 25 cm average major radius with a capacitor bank rise time of less than 2 ?sec. Two shaped toroidal coils with opposing currents and the poloidal compression coils have been designed to give a pressure balance equilibrium and establish the limit cycles. This device could be used to determine the physical significance of topological stability in plasma confinement.     

Investigation on the Hard X-ray Radiations of the IR-T1 Tokamak Plasma: Electric and Magnetic Perspectives

In this experiment, the effect of magnetohydrodynamic (MHD) fluctuations in the hard X-ray radiation from the IR-T1 tokamak plasma is investigated. To reach this goal, the main parameters of plasma such as plasma current and loop voltage are measured. Also, the rake and poloidal Langmuir probes are used to calculate the radial and poloidal electric fields. To detect the hard X-ray radiation, a NaI-scintillator detector is used. To study on the MHD fluctuations, an array of 12 Mirnov coils is used. The obtained data are analyzed by using the singular value decomposition (SVD) algorithm. The wavelet spectrum of the dominant principal components of Mirnov coils is drawn. The results of wavelet and SVD analysis show that the hard X-ray radiation is increased with increasing the fluctuations of the dominant principal components (at the same time). It is also shown that the rate of hard X-ray radiation emitted from the tokamak plasma increased with increasing the MHD fluctuations. The energy of the system is wasted and reduced by these radiations. This an increase in the particle pressure of the plasma.