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用TSC程序模拟了EAST装置等离子体放电的全过程。模拟中考虑了自举电流，并加入了离子回旋共振加热ICRH和快波电流驱动FWCD，得到了中心电子温度4.5keV、中心离子温度3.8keV、中心电子密度1.2×1020m–3的D形截面的等离子体。根据模拟结果对EAST装置进行了伏秒数分析，并研究了不同等离子体电流上升时间、有效电荷数Zeff对放电的影响。     

Tokamak中自举电流的剖面准直性

利用Harris模型,通过求解等离子体平衡方程,计算俘获粒子份额,分别对常规剪切和中心负剪切下tokamak中的自举电流的大小和剖面准直性进行了计算和分析.自举电流分布与等离子体平衡电流分布之间的剖面准直性可以通过调整等离子体的密度、温度和电流分布参数,以及描述等离子体形状的拉长度k和三角变形因子d来获得.中心负剪切位形有利于自举电流产生,并有好的剖面准直性.通过计算比较,分别在常规剪切位形下和中心负剪切位形下获得了一组优化的等离子体参数,在这组参数下,自举电流有较大的份额和好的剖面准直性 关键词： tokamak 自举电流 剖面准直性     

低混杂波电流驱动中原副边回路相互作用

本文用速度空间二维高速电子Ｆｏｋｋｅｒ－Ｐｌａｎｃｋ方程和原、副边回路方程组成的方程组描述托卡马克装置低浊经杂民流驱动宏观参数，研究了低混杂波电流驱动中的电场效应和原、副边回路相互作用应用该程序，初步模拟了ＨＬ－１装置的ＬＨＣＤ实验。     

托卡马克芯区Zeff测量和分析方法的研究

基于等离子体锯齿弛豫振荡的测量和研究,结合芯区电子功率平衡的分析,获得一种可靠的确定芯区平均有效离子电荷数Zeff的测量方法。该方法不受杂质的限制,测量条件十分容易满足。     

多道干涉仪测量托卡马克等离子体电子密度

一、引 言 在目前托卡马克等离子体诊断中,采用垂直多道干涉仪测量等离子体电子密度时,由于等离子体水平位移的影响,使得测量信号在沿大环半径方向的分布是非对称的。这就要求采用非对称的Abel变换来给出等离子体电子密度的空间分布。一种常用的非对称Abel变换是分离变量法,它的基本思想是把测量信号分成对称元和非对称的权函数,对称部分采用标准的Abel变换,变换的结果乘上非对称的权函数。这种方法实际上是假定了等离子体电子密度弦积分值的不对称性与局部的不对称性是相同的;而且对称部分采用标准的Abel变换,如Barr的方法,Bockasten方法等。这些方法是把等离子体看成一个个同心圆,没有考虑到等离子体水平位移引起的弦长变化。本文采用一种新的变换方法,这种方法的特点是考虑到等离子体位移的实际情况。     

非圆截面托卡马克等离子体边缘磁场的解析解及重构方法

本文用傅里叶及边缘泰勒展开的方法，导出了非圆截面托卡马克等离子体边缘磁面的解析解及其重构的方法。最后作为例子，给出了在ＴＥＸＴ－Ｕ装置中按实际参数重构的等离子体边缘磁面结构及重构的误差。     

HL-2A 装置可见光诊断方法初探

等离子体H_α 径向分布的测量是等离子体实验中的重要内容。在HL- 2A 实验中, 等离子体成像系统中的CCD 图片记录了等离子体辐射信息, 特别是H_α  线辐射信息( 在相机前加H_α 滤光片) 。因为光测量是辐射的线积分, 数据的处理和分析显得十分必要的。利用等离子体可见光成像系统, 采用Abel 变换, 得到H_α  在HL- 2A 装置中的径向分布, 提供了一项可靠的测量手段。     

EAST远红外激光干涉仪与密度爬升实验研究

利用EAST装置单道远红外HCN激光干涉仪测量了等离子体中心道(R=1.82m)线平均电子密度。通过充气加料连续提升主等离子体密度,首次在EAST装置上观察到偏滤器等离子体的三种不同状态：低再循环(偏滤器靶板处等离子体温度较高,密度较低),高再循环(偏滤器靶板处等离子体温度较低,密度较高)和脱靶(偏滤器靶板处等离子体温度和密度都很低)等离子体状态。分析了EAST偏滤器在这三种不同状态下的物理现象。     

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利用EAST装置单道远红外HCN激光干涉仪测量了等离子体中心道(R=1.82m)线平均电子密度。通过充气加料连续提升主等离子体密度,首次在EAST装置上观察到偏滤器等离子体的三种不同状态：低再循环(偏滤器靶板处等离子体温度较高,密度较低),高再循环(偏滤器靶板处等离子体温度较低,密度较高)和脱靶(偏滤器靶板处等离子体温度和密度都很低)等离子体状态。分析了EAST偏滤器在这三种不同状态下的物理现象。     

Eddy current effects on plasma equilibrium in a single-turn tokamak

In a fusion experiment based on the single-turn tokamak concept, the plasma is surrounded by a massive conducting structure composed of several layers of material with different resistivities. This conducting shell is located near the plasma edge and is magnetically coupled to the plasma column. The plasma magnetohydrodynamic (MHD) equilibrium is studied by neglecting the effect of structural induced currents. Eddy current effects are then analyzed. Poloidal uniformization of the poloidal field magnet current distribution required for plasma equilibrium is demonstrated. The possibility of continuous-limiter discharges in a single-turn tokamak configuration is pointed out. The significance of these results for the operation of a high-current tokamak experiment is discussed     

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.     

Magnetohydrodynamic equilibrium interpretation of low-aspect-ratio tokamak plasma

It is very important to calculate the equilibrium magnetic field configuration exactly and to estimate the parameters of the device needed for the equilibrium and stability in the design and operate the low-aspect-ratio tokamak, one of the axis-symmetrical torus devices which is widely used in the research of nuclear fusion using plasma. In the previous researches, there were the theoretical and numerical interpretation methods for the high-aspect-ratio tokamak plasma. But for the low-aspect-ratio tokamak plasma, numerical methods such as the finite element method and iteration method are usually used. In this research, we interpreted and examined the features of magnetohydrodynamic equilibrium of the low-aspect-ratio tokamak plasma in combination with the Green function method and finite element method which has high numerical value accuracy.     

Reinstallation of the COMPASS-D tokamak in IPP ASCR

The COMPASS-D tokamak, originally operated by UKAEA at Culham, UK, will be reinstalled at the Institute of Plasma Physics (IPP) AS CR. The COMPASS device was designed as a flexible tokamak in the 1980s mainly to explore the MHD physics. Its operation (with D-shaped vessel) began at the Culham Laboratory of the Association EURATOM/ UKAEA in 1992.The COMPASS-D tokamak will have the following unique features after putting in operation on IPP Prague. It will be the smallest tokamak with a clear H-mode and ITER-relevant geometry. ITER-relevant plasma conditions will be achieved by installation of two neutral beam injection systems (2 × 300 kW), enabling co-and counter-injections. Redeployment of the existing LH system (400 kW) is also envisaged. A comprehensive set of diagnostics focused mainly on the edge plasma will be installed.The scientific programme proposed for the COMPASS-D tokamak installed in IPP Prague will benefit from these unique features of COMPASS-D and consist of two main scientific projects, both highly relevant to ITER-Edge plasma physics (H-mode studies) and Wave-plasma interaction studies.The COMPASS-D tokamak will offer an important research potential as a small, flexible and low-cost facility with ITER-relevant geometry.     

Determination of Plasma Position Using Poloidal Flux Loops and Comparison With Magnetic Probe Measurement in IR-T1 Tokamak

In this paper, we present two magnetic techniques for the measurement of plasma position in IR-T1 tokamak: a poloidal flux loop and a magnetic probe method. In the first method, two flux loops were designed and installed toroidally on the outer surface of the IR-T1 tokamak, and then, displacement of the plasma column was measured from them. In addition, to compare the plasma position obtained using the flux loops, an array of four magnetic probes was designed, constructed, and installed on the outer surface of the IR-T1 tokamak, and plasma position was measured from them. Results were compared and found to be in good agreement with each other.      

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通过HCN信号的测量与处理来获得等离子体电子密度。采用硬件相位差计获取HT-7装置中原始HCN信号,根据对该信号特点的分析, 设计出信号去零漂、信号翻转、基于模糊逻辑的去噪处理算法。在HT-7装置实验中,使用该算法获得了较高精度的低噪音等离子体电子密度。     

Methods for reconstructing equilibrium plasma configurations in the Globus-M spherical tokamak

A magnetic diagnostics allowing one to reliably reconstruct equilibrium plasma configurations in a tokamak over a wide range of operating parameters is developed. The accuracy of determining the geometrical parameters and thermal energy of the tokamak plasma is analyzed in detail. The experimental data obtained in the Globus-M tokamak are processed the with help of the EFIT code. The influence of the plasma configuration on the intensity of the main impurity lines is investigated.     

Characteristic of Disruption on HL-2A Tokamak

The tokamak disruption is a dramatic event in which the plasma confinement is suddenly destroyed. Detailed experimental studies of disruptions have been made in many machines. During disruption, the plasma current and plasma thermal energy content collapse in an uncontrollable way, thereby applying mechanical forces and heat loads onto the vacuum vessel components. For that reason, the disruptions in a tokamak must be investigated and the physical processes leading to and occurring at the disruption need to be understood.     

HL—1装置偏压电极类H模能量约束时间研究

能量约束时间是衡量环流器等离子体约束性能的重要参数。分析表明，在加偏压电Ｌ模过渡到类Ｈ模的过程中，如果等离子体的辐射损失功率与总损失功率之比显著变化，则扣除辐射损失的能量约束时间的增量是一种更好的衡量约束得到改善的尺度。在这种考虑之下，我们讨论ＨＬ－１等离子体偏压电极Ｌ模－类Ｈ模过的能量约束及电子热传导特性。     

Current clamp at zero level in JT-60U current hole plasmas

It is found that no current is driven in a central region of a tokamak plasma once the central current density becomes nearly zero ("current hole"), in spite of high electric conductivity, at the current drive by a toroidal electric field and a radio-frequency wave in experiments on the JT-60U tokamak. This is a new, stiff, self-organized structure of a magnetic field in an axisymmetric toroidal plasma.     

HL-1M装置等离子体约束时间

强调了托卡马克等离子体的约束时间是等离子体粒子损失或热能损失的特征时间,评述了有关物理概念和实验现象分析了方面的问题,并应用于HL－1M装置等离子体。