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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。     

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。     

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。     

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     

HL-2M等离子体垂直不稳定性的反馈控制研究

使用无源稳定导体和主动控制线圈来控制HL-2M等离子体的垂直不稳定性。计算了等离子体垂直不稳定性增长时间,构建了等离子体垂直不稳定性控制的线性模型,然后用MATLAB对采用PID算法的垂直不稳定性控制系统进行了模拟仿真。结果表明,无源稳定导体和主动控制线圈能够快速稳定等离子体的垂直不稳定性运动,这也表明设计是可行的。     

HL—1装置等离子体位置反馈控制

本文简述了用位移做反馈控制信号,用非线性二位调节器组合前馈环节,控制厚钢壳(厚度d=5cm)外的垂直场和水平场线圈电流,在HL-1装置上实现了等离子体位置反馈控制的情况。当铁芯不饱和时,在水平和垂直两个方向上均可控制等离子体位置在±2mm之内。     

强磁场下ITER垂直稳定性线圈的电磁性能研究

垂直稳定性线圈是ITER托卡马克装置中用以维持等离子体垂直稳定性的核心部件,然而该线圈处于强磁环境下,一旦线圈励磁运行以后,线圈电流与外部背景场交互作用激发的电磁载荷将会对线圈产生强烈的电磁冲击.为了分析稳态和瞬态电磁载荷对线圈结构的影响,首先采用电流丝等效模型并基于椭圆积分计算了背景场线圈和等离子体电流的磁场,以及极端运行状态下背景场所诱发的电磁力.在电磁计算的基础上,创建循环对称磁-结构耦合分析模型并采用间接耦合法,将电磁力密度以载荷边界条件插值到结构分析模型,计算评估线圈在稳态电磁载荷下的力学性能.此外,针对脉冲电流引起的电磁疲劳,采用ASME主应力方向恒定的疲劳设计规范,结合Goodman修正法并以无限次循环条件下材料应力强度为评定标准,对线圈各组件的应力进行了评定,结果表明线圈各组件均有足够的安全裕度.该分析方法可为托卡马克其它磁体线圈的电磁性能评估提供参考.     

三维数值模拟射频热等离子体的物理场分布

射频热等离子体的产生包含了丰富、复杂的物理场分布,正确认识这些物理场分布对射频热等离子体在工业领域的应用有指导作用.本文建立了三维射频热等离子体的热-电-磁-流动强耦合数学物理模型,考虑了感应线圈的真实螺线管结构,开发了用于计算三维射频热等离子体复杂电磁场的C++程序代码,计算了射频热等离子体重要物理场,如温度场、流场和电磁场的分布情况.结果表明,射频热等离子体各物理场分布具有三维非对称效应,感应线圈结构对温度场和流场的空间分布有重要影响.研究结果对优化、控制射频热等离子体的实际应用过程有重要的指导意义.     

HL－1M装置等离子体的平衡和控制

ＨＬ－１Ｍ是ＨＬ－１的改进装置，去掉铜壳后，对极向场线圈和电源都做了必要的修改。此外，还增加了四个快垂直场线圈，并采用可控环流四象限晶闸管交流器供电。对装置调试和运行结果表明．反馈控制系统能很好地维持等离子体平衡。放电一直持续到铁芯变压器的伏秒数用完为止，最长达１０４０ｍｓ。在多数放电的平顶段，等离子体位置都被控制在±４ｍｍ之内．     

托卡马克装置中等离子体环向旋转对三维响应场的影响

本文利用MARS-F程序,数值研究了HL-2M托卡马克装置高比压等离子体中,环向旋转对外加共振磁扰动场的响应特性的影响.研究发现,等离子体响应显著改变共振磁扰动的谱分布,并影响等离子体内部共振磁扰动场与共振磁扰动线圈电流相位差的依赖关系,从而改变有理面处径向扰动场的幅值.当边界旋转频率较小时,在最外有理面处,等离子体响应对外加共振磁扰动场有明显的放大效应.通常,边缘局域模的控制效果依赖于最外有理面处共振磁扰动场的幅度,因此可通过控制旋转剖面实现对共振磁扰动场的调控,进而优化边缘局域模的控制方案.     

Comparison of 12 Quadrature Birdcage Coils with Different Leg Shapes at 9.4 T

The purpose of this study was to analyse the relationship between the radio frequency (RF) coil performance and conductor surface shape for ultra-high field (UHF) magnetic resonance imaging. Twelve different leg-shaped quadrature birdcage coils were modeled and built, e.g., 4 mm-width-leg conventional birdcage coil, 7 mm-width-leg conventional birdcage, 10 mm-width-leg conventional birdcage coil, 13 mm-width-leg conventional birdcage coil, inside arc-shape-leg birdcage coil, outward arc-shape-leg birdcage coil, inside right angle-shape-leg birdcage coil, outward right angle-shape-leg birdcage coil, vertical 4 mm-width-leg vertical birdcage, 6 mm-width-leg vertical birdcage, 8 mm-width-leg vertical birdcage and 10 mm-width-leg vertical birdcage. Studies were carried out in both electromagnetic simulations with finite element method as well as in vitro saline phantom experiments at 9.4 T. Both the results of simulation and experiment showed that conventional birdcage coil produces the highest signal-to-noise ratio (SNR) while the vertical birdcage coil produces the most homogeneous RF magnetic (B ₁) field at UHF. For conventional birdcage coils, as well as the vertical birdcage coils, only the proper width of legs results in the best performance (e.g., B ₁ homogeneous and SNR). For vertical birdcage coils, the wider the leg size, the higher RF magnetic (B ₁) field intensity distribution.     

The dispersion and matching characteristics of the helicalresonator plasma source

The dispersion characteristics and the fields of a helical resonator are obtained. The coil is approximated by a helical current layer with infinite conductivity along the current direction (a “sheath helix”). The plasma column is modeled as a cylindrical dielectric in which the dielectric losses can be neglected in determining the propagation properties of the resonator. Assuming the plasma losses are known, the model can be used to study the matching of the helical resonator to an external power source which is connected to the helix by a tap. The resonator is modeled as a parallel connection of two transmission line segments on each side of the tap position. We determine the efficiency of power transfer to the resonator as a function of the tap position driving frequency, and plasma loading. We find that whereas for a small plasma loading it is possible to achieve perfect matching, there exists a critical value of plasma loading beyond which a perfect match is no longer possible     

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。     

4T split TEM volume head and knee coils for improved sensitivity and patient accessibility

Split RF coils offer improved patient access by eliminating the need for the coil to be slid over the region of interest. For unshielded birdcage coils, the presence of end ring currents necessitates a direct electrical connection between two halves of the coil. For high-field (>3T) shielded birdcage coils, both the shield and the coil must be split and reliably connected electrically. This problem can be circumvented by the use of split TEM volume coils. Since the elements of a TEM coil are coupled inductively, no direct electrical connection between the halves is necessary. In this work we demonstrate that the effects of splitting the shield for head and knee TEMs can be compensated for, and performance retained. For the knee, the improved access allowed the coil diameter to be reduced, enhancing the sensitivity by 15-20%.     

Negative edge plasma currents in the SINP tokamak

A tokamak plasma discharge having an increase in duration accompanied with enhanced runaway electron flux has been experimentally studied in this paper. The discharges have been obtained by controlling the applied vertical magnetic field (B_v^applB_{\rm{v}}^{\rm{appl}}) to below a critical value. Such discharges have been observed to have ‘negative edge plasma currents’, detected using an internal Rogowskii coil (IRC). We have tried to correlate the runaway behaviour with the negative edge plasma currents and have explained that these observations are a result of beam plasma instabilities.     

Eliminating islands in high-pressure free-boundary stellarator magnetohydrodynamic equilibrium solutions

Magnetic islands in free-boundary stellarator equilibria are suppressed using a procedure that iterates the plasma equilibrium equations and, at each iteration, adjusts the coil geometry to cancel resonant fields produced by the plasma. The coils are constrained to satisfy certain measures of engineering acceptability and the plasma is constrained to ensure kink stability. As the iterations continue, the coil geometry and the plasma simultaneously converge to an equilibrium in which the island content is negligible. The method is applied with success to a candidate plasma and coil design for the National Compact Stellarator Experiment [Phys. Plasmas 8, 2083 (2001)]].     

A Study of Magnetic Diverter Properties in the WISTOR-U Ultimate Torsatron

The WISTOR-U ultimate torsatron design was modeled to determine the effect of various vertical coil configurations particularly with respect to divertor properties. The separatrix and divertor behavior were studied using a modified version of a previously developed code. The divertor properties were studied for various vertical field coil configurations and strengths. The possibility of using structural members for vertical coils and the resultant divertor action around the structural members was investigated. A reduction in area inside the separatrix was observed with deviations from the pure ultimate torsatron base case.     

Experimental Studies on the Sustainment of Splieromak Plasmas by an Inductive Drive of the Toroidal Current

Sustainment of spheromak plasmas produced in an external equilibrium field has been demonstrated with a center current transformer (ohmic heating (OH) coil) which is used to inductively drive the toroidal current of the plasma. The OH coil is covered by a cylindrical metal liner. It provides the stability against the tilt and shift motions of spheromaks at the expense of the simple connection of its geometry. Since the spheromak is characterized by the elimination of external toroidal fields in association with nonconservation of a toroidal flux during magnetic relaxation, the metal liner was made electrically disconnected from the main vacuum vessel (spheromak mode). In the experiments, existense of the dynamo effect, meaning automatic generation of toroidal flux similar to that of a reversed field pinch (RFP), is observed. Measured MHD activity consists of multihelicity helical modes with toroidal mode numbers N = 1-3. In order to investigate the difference between spheromaks and RFP's in the MHD activity during sustainment, experiments have also been made with the metal liner of the OH coil connected with the vessel (RFP mode). The dynamics of the MHD activities observed are compared with those obtained from the three-dimensional MHD simulations by Katayama and Katsurai [18], and their implication in the dynamo effect is discussed.