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。     

Energy Selecting Action of Limiters on Particle Fluxes Penetrating into the SOL-Plasma

A simple model of the penetration of particle effluxes from the core plasma into the SOL-plasma of tokamaks is proposed. The assumptions made are free streaming of particles parallel to the magnetic field and anomalous particle transport perpendicular to the toroidal field with a constant radial velocity. The model has been proved for measured particle fluxes of Li which was injected into the core plasma of the tokamak T-10. The dependence of the Li-particle flux on the minor radius as well as toroidal asymmetries in the SOL-plasma can be explained by the results of the calculations.     

Toroidicity and shape dependence of peeling mode growth rates in axisymmetric toroidal plasmas

The growth rate of the peeling mode instability with large toroidal mode number is calculated for general axisymmetric toroidal plasmas, including tokamaks and the spherical torus(ST) equilibia by using formalism presented by Connor et al.Analytic equilibia with non-zero edge current density and quasi-uniform current profiles are assumed. It is found that in sharp D-shape tokamak plasma, the derivative of the safety factor with respect to the poloidal flux becomes very large,making the perturbed poloidal motion very large, in turn making a significant reduction of the growth rate of the peeling mode, similar to the X-point effect in diverted plasma. The large aspect ratio effect is also studied, which reduces the growth rate further.     

Estimation of 14 MeV neutron rate from triton burn-up in future W7-X deuterium plasma campaigns

Fast ion confinement is of major importance for the ignition of a burning fusion plasma. In future deuterium plasma campaigns of the Wendelstein 7-X stellarator, W7-X, the amount of triton burn-up is one possible measure for fast ion confinement. A well-established technique to observe triton burn-up is the 14 MeV neutron rate. In this paper, it is estimated whether an existing scintillating fibre neutron detector is also suited to measure triton burn-up in W7-X with sufficient accuracy. An estimation is presented, which can be applied to any tokamak or stellarator design and is one-dimensional in the minor radius. The inputs are profiles of density, temperature, and differential volume element as well as the triton slowing-down time. The estimation calculates the thermal deuteron fusion rate and the associated deuteron-triton fusion rate; thus, the triton burn-up generated 14 MeV neutron rate. It neither takes triton diffusion nor explicit losses into account. This thermally generated fusion rate is compared to the neutral beam injection heating induced beam-plasma fusion rate.     

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。     

D-~3He商用堆工作物理参数的空间分析

由于D-~3He聚变堆不要求厚的包层和大的内侧空间,中子壁负载小得多,第一壁几乎是永久性的,维修更换问题减轻。采用低径比有利于电流驱动和较大的等离子体变形和好的垂直稳定性,以便得到大的等离子体电流,因而得到高的β和好的约束。本文从目前普遍采用的堆芯等离子体物理规律为基础,编制了DHE3TOK程序。研究了以D-~3He为燃料,低径比A=2.5工作在第一稳定区的商用堆特点。对它的工作物理参数空间范围进行广泛调查的基础上,提出了需要研究和发展的关键工程技术问题。     

Operation of a tokamak reactor in the radiative improved mode

The operation of a nuclear fusion reactor has been simulated within a model based on experimental results obtained at the TEXTOR-94 tokamak and other facilities in which quasistationary regimes were achieved with long confinement times, high densities, and absence of the edge-localized mode. The radiative improved mode of confinement studied in detail at the TEXTOR-94 tokamak is the most interesting such regime. One of the most important problems of modern tokamaks is the problem of a very high thermal load on a divertor (or a limiter). This problem is quite easily solved in the radiative improved mode. Since a significant fraction of the thermal energy is reemitted by an impurity, the thermal loading is significantly reduced. As the energy confinement time τ_E at high densities in the indicated mode is significantly larger than the time predicted by the scaling of ITERH-98P(y, 2), ignition can be achieved in a facility much smaller than the ITER facility at plasma temperatures below 20 keV. The revealed decrease in the degradation of the confinement time τ_E with an increase in the introduced power has been analyzed.     

Status of Stellarator Experiments?1981

During the last years growing interest in stellarators as toroidal magnetic confinement devices becomes evident. In spite of many restrictions, e.g., small minor radius a ~0.1 m, plasma parameters comparable to those in tokamaks have been achieved. The deleterious effect of the plasma current on stability, equilibrium, and transport could be clarified. The progress of heating by the use of neutral injection and RF, allows to substitute ohmic heating. In the currentless phase significant improvement of the confinement at reasonable parameters has been observed. The substantial reduction of low-frequency and MHD fluctuations seems related to a decreased electron heat conduction.     

Effect of subsonic toroidal flows on plasma poloidal rotation and ion heat conductivity. In collisional plasmas of elongated tokamaks

The ion poloidal rotation and heat conductivity in collisional plasmas of axially-symmetric tokamaks with elongated cross-sections and with subsonic toroidal plasma flows are considered. It is shown that subsonic toroidal plasma flows, induced by neutral beam injection or radio frequency waves, can strongly affect the poloidal plasma velocity and ion heat conductivity in collisional plasmas of tokamaks. The transport coefficients also depend on the tokamak ellipticity parameter which, in combination with the Mach number, allows to operate transport processes at smaller values of the toroidal Mach number. The importance of taking into account the ion-electron heat exchange and electron temperature toroidal perturbations to find ion temperature toroidal perturbations is demonstrated. This work was partially supported by the State University of Rio de Janeiro (UERJ) and the Rio de Janeiro Research Foundation (FAPERJ).     

Laplace series expansion of the internal potential of a homogeneous circular torus

The internal potential of a homogeneous circular torus first is represented by a series expansion in spherical functions (Laplace series). Exact analytical formulas for the coefficients of this series are derived and it is shown that they can be expressed through the standard Gauss hypergeometric function depending only on the geometric parameter of the torus. Convergence of the series is proved and the radius of convergence is determined. The relation of the radius with the torus geometrical parameter is found. A special spherical shell, where the problem of expansion of the torus potential should be solved in additional investigations, is detected.     

托卡马克中等离子体柱的位移稳定性

本文考虑托卡马克中纵场的作用以及等离子体柱和与之有互感的外迴路的相互作用,在圆截面,瘦环假设下,得到了自由运动时相对于沿水平方向、垂直方向的位移,以及小半径变化的一般稳定条件。由于强纵场的影响,一般来说,等离子体柱的小半径总是随着大半径的变化而改变,以维持通过等离子体角向电流迴路的磁通不变。对于这种扰动,当考虑到外迴路电流变化以后,稳定区域超过公式0关键词：     

Design study of 3-D tomography diagnostics for spherical tokamaks

A novel three-dimensional (3-D) tomography software has been developed for spherical tokamak (ST) plasmas with thin central coils. It enables us to measure whole 3-D annular profiles of ST plasma emissivity (line spectrum etc.) using limited number of detectors located on a sphere outside of its separatrix. The numbers and positions of detectors have been optimized for STs that contain low-n toroidal modes from n = 1 to 5. Mathematical method allowing for the removal of the thin central coil from the region of reconstruction and to reconstruct STs with single and double axes are also discussed. The shadow effect of center coil region was found to limit the reconstruction accuracy, indicating that the low-aspect ratio torus (ST) is essential to this new 3-D tomography diagnostics. Received 17 August 2000 and Received in final form 16 November 2000     

Analytic description of tokamak equilibrium sustained by high fraction bootstrap current

By using an expansion technique based on the tokamak ordering, the Grad－Shafranov equation is analytically solved for tokamak equilibrium sustained by a full bootstrap current. This approximate approach is suitable for equilibria with non-circular cross-sections of modest ellipse and triangular deformations. As an input parameter set, the plasma pressure profile and the electron and the ion temperature profiles can be selected as arbitrary functions of the averaged minor radius. Equilibrium properties of this plasma are discussed.     

Dynamical Effects in Confined Plasma Turbulence

Plasma turbulence at the edge of tokamaks is an issue of major importance in the study of the anomalous transport of particles and energy. Although the behavior of a turbulent plasma seems intractable, it turns out that many of its aspects can be described by low-dimensional non-integrable dynamical models. In this paper, we consider a number of dynamical effects occurring in tokamak plasma edge—in particular the role of internal transport barriers. Furthermore, we present experimental results on turbulent-driven transport for two machines—the Brazilian TCABR tokamak and University of Texas’ Helimak—that can be explained by those theoretical models.     

环径比为2的托卡马克反应堆中氘氚燃烧等离子体特性

本文对环径比为2（A=2）的托卡马克反应堆中氘氚燃烧等离子体的特性，利用一类简单的平衡位形，即Solov’ev位形，进行了研究。作为基准，采用了Troyon比压极限和Greenwald密度极限，以及 ITER的H-模约束定标关系。我们发现，除了增大拉长比外，大的三角变形也有利于提高比压值及获得高的聚变功率输出。与 ITER设计相比，对相同的总等离子体电流而言，A=2的反应堆因边缘安全因子很大，具有可以避免破裂不稳定性的优点。由于减小环径比时环内侧的可利用空间相应减小，这类反应堆的一个主要缺点是环向磁场强度也会减小，从而降低聚变堆品质。实际研究工作中要对这些方面做出某种折衷。另外，虽然由于我们采用了简单的位形，我们不能对自举电流的准直性进行优化，但是通过本研究仍然可以对中等环径比反应堆系统（目前是研究工作的空白）的特点获得了解。     

受控热核聚变研究及其在我国HT-7超导托卡马克上的最新进展

受控热核聚变研究在发展了50年后，在2003年进入了一个新的阶段，国际热核聚变实验堆项目将最后选址开工。科学工作者们希望这个在新世纪初开始的项目能成为彻底解决能源问题的发端，中国科学院等离子体物理研究所最近在受控热核聚变科研上取得了长足进展，朝着稳态高温等离子体的方向又跨出了坚实的一步．其HT-7超导托卡马克高温等离子体的持续时间达到63．95s，成为目前世界上能开展分钟级高温等离子体放电的仅有的两个超导托卡马克装置之一。文章主要介绍了中国科学院等离子体物理研究所最新的聚变实验研究进展。     

托卡马克极向场线圈的优化方法

托卡马克等离子体的三角形变和拉长比对约束和磁流体稳定性有很强的影响，因此在托卡马克装置极向场设计中，在基于物理和工程考虑所预先选定的等离子体平衡位形几何参数下，如何优化确定外部极向场线圈位置和电流，是一个具有重要实际意义的研究课题.为优化确定托卡马克极向场线圈，给出了一个有效的多变量平衡优化方法，能以事先规定的等离子体平衡位形的一些几何参数为目标函数，优化确定极向场线圈位置和电流.并应用它于HT-7U平衡位形计算，得到了所需的结果. 关键词： 等离子体平衡 极向场线圈 优化     

低混杂波高N‖分量对EAST等离子体电流驱动的影响

EAST等离子体高约束模运行条件下，在等离子体边缘区域观测到明显的等离子体电流带.在EAST托卡马克装置非圆截面平衡位形下，使用射线追踪方法研究低混杂波高平行折射率N_‖分量对电流驱动的影响.结果表明：当-8≤N_‖≤-6时，平行折射率分量能够在小半径（0.7 < r/a < 1）区域驱动kA量级的等离子体电流.对于具有台基区、等离子体边缘温度更高的电子温度剖面，驱动电流的位置r/a>0.9.低混杂波朗道阻尼的理论分析与数值模拟结果一致.另外，高N_‖低混杂波在等离子体边缘的功率沉积和电流驱动与电子温度分布和发射谱分布相关.     

Electronic Circuits for the Generation and Transfer of High-Power Pulses in Nuclear Fusion Installations

The toroidal confinement and the ohmic heating of a thermonuclear plasma as used in tokamak experiments can be achieved by strong magnetic fields and by a heavy direct current flowing parallel to the axis of the plasma torus. This current can be induced with the aid of a transformer for which the plasma torus acts as the secondary winding. The high loop voltage required for start-up is obtained by commutation of the primary current to a resistor. This commutation can be achieved by a mechanical switch or by electronic circuits. New electronic commutation circuits are described which enable very heavy direct currents to be switched without contacts. The operation of the new circuits was checked by computer calculations. Experiments have not yet been performed.     

环流器等离子体高n气球模的第二稳定区

本文讨论了圆截面高比压等离子体关于高n气球模的稳定性,在高比压情况下,相应的极向磁场对气球模有相当强的驱动作用,从而严重影响了第二稳定区的结构,我们详细计算了不同剪切、不同压强梯度及不同极向场参数对气球模本征函数和本征频率的影响,这些结果比较完善地反映了圆截面环流器中理想磁流体气球模理论所预示的主要结论。 关键词：