自持稳态托卡马克等离子体

本文介绍了国家球形环实验（ＮＳＴＸ）和球形环体积中子源（ＳＴ－ＶＮＳ）的自持稳态的托卡马克平衡,证明了压力梯度驱动电流的量与在这些装置中设计的电流条件相当,前者包括自举电流和抗磁电流。也说明了通过增加等离子体压力来启动的可能的程序。     

在低混杂波频率范围大功率微波辅助下实现HT-7托卡马克低环电压启动

在HT－7托卡马克放电实验中，使用N∥谱实时可调，频率在等离子体低混杂波频率范围，功率达150－200kW的大功率微波，辅助托卡马克装置的加热场进行了等离子体放电。在低环电压下实现了托卡马克的成功启动。使用这种方法使HT－7装置的启动环电压由通常的20V左右下降到5V以下。这相当于实现了放电启动的最大加热场副边场强由2.5-3V/m降低到0.6V/m左右。实验还观察到，在这种微波辅助低环电压启动条件下，气体电离过程和等离子体电流建立过程中的伏秒数的消耗，初始等离子体的辐射等一系列不同于单纯欧姆放电过程的物理现象。     

KSTAR托卡马克设计

韩国超导托卡马克（KSTAR）探索性研究项目是韩国国家聚变计划的重点项目，其主要工作是研制能稳态运行的先进超导托卡马克，为具有吸引力的聚变堆建立科学技术基础。该托卡马克的主要参数是，大半径1.8m，小半径0.5m，环向场3．5T，环向场3．5T，等离子体电流2MA，并具有强变形的等离子体横截面和双零偏滤器。最初由极向磁系统提供的脉冲长度为20s，但是通过非感应电流驱动脉冲长度可以增大到300s。等离子体加热和电流驱动系统包括中性束，离子回旋波，低杂波和电子回旋波，它们都可用于灵活的剖面控制。全套诊断设备计划用于等离子体控制和特性计算以及对物理学的了解。该项目已经完成概念设计阶段，进入到工程设计阶段。首次产生等离子体的日期确定在2002年。     

托卡马克逃逸电子输运系数的研究

1 对托卡马克逃逸电子的一般介绍 近期,人们对托卡马克中逃逸电子输运的研究十分重视与活跃。从欧姆加热与辅助加热的等离子体中,逃逸约束时间具有反常行为,这与内部的磁场涨落相关;我们利用逃逸电子作为一种试验粒子来诊断托卡马克内部的磁涨落。从约束的环向等离子体中观察到粒子与能量的损失率远比由新纪典输运理论所预言的要大得多,此状态取决于平均等离子体参数与库仑碰撞。若干类型的等离子体湍流提供了附加的等离子体输运。等离子体参数的涨落将通过静电涨落或磁涨落引起输运。该涨落所驱动的径向粒子通量为(1)其中,是扰动角向电场,是扰动电子密度。是     

球形托卡马克装置SUNIST的建立

2002年建成的SUNIST球形托卡马克装置是中国第1台低环径比(R／α≈1．3)的托卡马克装置，具有氟橡胶交叉密封和电绝缘的结构。该装置主要设计参数为：大半径0．3111，小半径0．23m，自然拉长比1.6，截面中心磁场0．15T和等离子体电流50kA。在初步的欧姆加热实验中已得到了等离子体电流为52kA的可重复放电。     

托卡马克中离子回旋频率范围内快波电流驱动的研究

在现在和未来的先进托卡马克装置中,离子回旋频率范围内的快波电流驱动都是一种不可或缺的非感应电流驱动方式。本文利用全波方程得到波电场和磁场分布,代入扩散系数并求解Fokker-Planck方程,采用自行编写的程序对快波在等离子体中的电子吸收功率和驱动电流等物理量进行数值求解。初步研究了快波频率和等离子体温度对电子吸收功率及电流驱动效率的影响,结果显示快波能传播到托卡马克等离子体的中心区域并被电子吸收;快波频率对电子吸收快波功率有显著的影响,在70～100 MHz波频率范围内电子对快波吸收效果较好,能有效驱动电流;另外在该频段内随着等离子体温度的升高,电子吸收效果变差。     

离子回旋频率范围的加热和电流驱动状态

在离子回旋频率范围（ＩＣＲＦ）发射的快声波是托卡马克等离子体的主要加热方案之一，这个加热和电流驱动方案可能有很大的灵活性，既可利用各种波－粒子相互作用机制，也可利用快波能经受到的模转换，如设计适当，ＩＣＲＦ系统能用于离子或电子加热的电流驱动，而且还有定域沉积分布的优点，几乎的有现在的托卡马克现都使用ＩＣＲＦ加热和电流驱动，而且正在为ＩＴＥＲ研究ＩＣＲＦ系统。     

球形环研究概述

球形环是一种先进的磁约束核聚变研究装置 ,它保留了传统托卡马克装置的许多特点 ,比如说封闭磁面系统 ,中等旋转变换磁力线和好的等离子体约束性能等 ,同时又有低环径比 ,低纵向磁场 ,从而具有结构紧凑 ,造价相对低廉等优点 ,是有潜力的磁约束核聚变途径之一。概述了球形环的装置、实验、理论等方面的研究现状和有待探索的问题     

射频波电流驱动抑制双撕裂模不稳定性研究

采用不可压缩磁流体模型,在圆柱位形下研究了射频波电流驱动对双撕裂模不稳定性的影响。结果表明,托卡马克装置中沉积在有理面上的同向驱动电流能有效减缓3/1双撕裂模的发展。电流驱动幅值约占等离子体电流初始值的4%、驱动电流沉积宽度约为小环半径的7%时,抑制效果较好。此外,研究表明,在双撕裂模进入快速增长阶段前加入外部驱动电流,才能有效抑制撕裂模的磁爆发。     

卡勒姆的球形托卡马克规划

球形托卡马克（ＳＴ）是常规托卡马克中环径比最小的托卡马克,并且似乎在一种较简单的装置中具有吸引人的物理特性。ＳＴＡＲＴ（小型紧凑环径比托卡马克）实验在世界上首次证明了ＳＴ位形中热等离子体的特性。它从１９９１年１月至１９９８年３月在卡勒姆运行,获得了等离子体电流达３００ｋＡ,脉冲持续时间为￣５０ｍｓ。它的下一代装置ＭＡＳＴ即将峻工,是一个特建的高真空装置,设计的等离子体体积增加１０倍,其等离子体电流     

Initial Plasma Startup Test on SUNIST Spherical Tokamak

The goal of the Sino-United Spherical Tokamak (SUNIST) at Tsinghua University is to extend the understanding of toroidal plasma physics at a low aspect ratio (R/a ≈ 1.3) and to demonstrate a maintainable target plasma by non-inductive startup. The SUNIST device isdesigned to operate with up to 13 kA of ohmic heating field current, and to 0.15 T of toroidal field at 10 kA of discharge current. All of the poloidal fields can provide 30 mVs of Volt-seconds transformer. Experimental results of plasma startup show that SUNIST has remarkable characteristics of high ramp rate (dIp/dt ≈ 50 MA/s ), high normalized current IN of about 2.8 (IN = Ip/αBT)，and high-efficiency (Ip/IROD ≈ 0.4) production of plasma current while operating at a low toroidal field. Major disruption phenomena have not been observed from magnetic diagnostics of all testing shots. Initial discharges with 52 kA of plasma current (exceeding the designed value of 50 kA),2 ms of pulse length and 50 MA/s of ramp rate have been achieved easily with pre-ionized filament.     

SUNIST Microwave Power System

Experiments on the start-up and formation of spherical tokamak plasmas by electroncy clotron heating alone without ohmic heating and electrode discharge assisted electron cyclotron wave current start-up will be carried out on the SUNIST (Sino United Spherical Tokamak) device.The 2.45 GHz/100 kW/30 ms microwave power system and 1000 V/50 A power supply for electrode discharge are ready for experiments with non-inductive current drive.     

Development of KSTAR in-vessel components and heating systems

In-vessel components of the Korea Superconducting Tokamak Advanced Research (KSTAR) were developed for 2010 campaign to provide a crucial circumstance for achieving the strongly shaped and diverted plasma. Moreover, the in-vessel components such as limiter, divertor, passive stabilizer, in-vessel control coil (IVCC) system demonstrated good performances satisfying the original design concepts. In addition to the plasma facing components and the IVCC, in-vessel cryo-pump (IVCP) system was also installed to leverage divertor operation. Besides the in-vessel components, there have been substantial progresses in development of the heating and current drive system. The KSTAR heating and current drive system includes all kinds of the major heating systems such as neutral beam injection (NBI), ion cyclotron range of frequency (ICRF), electron cyclotron resonance heating and current drive (ECH and ECCD), lower hybrid current drive (LHCD) systems. As an initial stage for full equipment of the heating systems to total power of 26 MW, several key systems such as 1st NBI (called NBI-1), ICRF, and ECH-assisted startup system successfully demonstrated their excellent feasibilities in the design and performances for dedication to the 2010 campaign.     

Physics design requirements for the National Spherical Torus Experiment liquid lithium divertor

Recent National Spherical Tokamak Experiment (NSTX) high-power divertor experiments have shown significant and recurring benefits of solid lithium coatings on plasma facing components (PFCs) to the performance of divertor plasmas in both L- and H-mode confinement regimes heated by high-power neutral beams. The next step in this work is installation of a liquid lithium divertor (LLD) to achieve density control for inductionless current drive capability (e.g., about a 15–25% n_e decrease from present highest non-inductionless fraction discharges which often evolve toward the density limit, n_e/n_GW  1), to enable n_e scan capability (×2) in the H-mode, to test the ability to operate at significantly lower density (e.g., n_e/n_GW = 0.25), for future reactor designs based on the Spherical Tokamak, and eventually to investigate high heat-flux power handling (10 MW/m²) with long pulse discharges (>1.5 s). The first step (LLD-1) physics design encompasses the desired plasma requirements, the experimental capabilities and conditions, power handling, radial location, pumping capability, operating temperature, lithium filling, MHD forces, and diagnostics for control and characterization.     

The Influence of Neutral Beam Injection on the Heating and Current Drive with Electron Cyclotron Wave on EAST

Both neutral beam injection(NBI) and electron cyclotron resonance heating(ECRH) have been applied on the Experimental Advanced Superconducting Tokamak(EAST)in the 2015 campaign. In order to achieve more effective heating and current drive, the effects of NBI on the heating and current drive with electron cyclotron wave(ECW) are analyzed utilizing the code TORAY and experimental data in the shot #54411 and #54417. According to the experimental and simulated results, for the heating with ECW, NBI can improve the heating efficiency and move the power deposition place towards the inside of the plasma. On the other hand, for the electron cyclotron current drive(ECCD), NBI can also improve the efficiency of ECCD and move the place of ECCD inward. These results will be valuable for the center heating, the achievement of fully non-inductive current drive operation and the suppression of magnetohydrodynamic(MHD) instabilities with ECW on EAST or ITER with many auxiliary heating methods.     

Lower hybrid heating and current drive system for a tokamak reactor

A lower hybrid heating system has been designed for heating a tokamak reactor to ignition and for sustaining steady-state operation by driving the toroidal plasma current. The power spectrum from an active/passive waveguide grill is computed, and the resulting equilibrium current density profile is computed from a full electromagnetic WKB analysis of wave propagation in a cylinder. The corresponding toroidal current profile is a low-current equilibrium which is stable to various ideal modes at an economically acceptable beta. The electronic circuitry is designed to minimize the electric power required for current drive, and the resulting design appears to provide reliable operation in a reactor environment. The same system can drive current during reactor startup if some of the waveguides are modified slightly. A typical sequence of startup equilibria is calculated.     

Analysis of the Gas Puffing Performance for Improving the Repeatability of Ohmic Discharges in the SUNIST Spherical Tokamak

The gas puffing performance plays a key role in repeatable discharges in the Sino- UNited Spherical Tokamak （SUNIST） experiments. In this paper, temporal evolution of the gas pressure in the vacuum vessel and the dependence of the repeatability of plasma discharges on different timing arrangements between the gas puffing pulse and the Ohmic field have been experimentally investigated. The results show that, after a fast rising phase, the gas pressure becomes quasi-stationary. In the regime of the discharges being started up when the gas pressure has already reached the quasi-stationary state for about 37 ms, an improved repeatability of the plasma discharges is achieved.     

EAST-ICRH系统高功率同轴隔直器的研制

EAST超导托卡马克是我国开展受控核聚变研究的新一代实验装置,离子回旋波加热(ICRH)是在该装置中加热等离子体的重要手段之一。离子回旋加热系统中的发射机和天线两者各有一个地电平,两个地电平会存在严重的相互干扰。高功率同轴隔直器的作用是用来隔断两者之间的直流通路,从而把两端的地电平分开,保证发射机和天线系统的正常运行。论文介绍了隔直器的原理、工程结构和设计指标,分析了S参数、端口驻波比的计算和仿真,最后给出了实际的隔直器参数曲线。     

Preliminary Experiment of Non-Inductive Plasma Current Startup in SUNIST Spherical Tokamak

The non-inductive plasma current startup is an important motivation in SUNIST spherical tokamak. In the recent experiment, the magnetron microwave system of 100 kW and 2.45 GHz has been used to the ECR plasma current startup. Besides the toroidal field, a vertical field was applied to generate preliminary toroidal plasma current without the action of the central solenoid. As the evidence of plasma current startup with the effect of vertical field drift, the direction of plasma current is changed when the direction of vertical field changes during the ECR plasma current startup discharge. We also observed a maximum plasma current by scanning vertical field in both directions. Additionally, we used electrode discharge to assist the ECR plasma current startup.     

Microwave preionization and electron cyclotron resonance plasma current startup in the EXL-50 spherical tokamak

Preionization has been widely employed to create initial plasma and help the toroidal plasma current formation. This research focuses on implementing a simple, economical and practical electron cyclotron resonance (ECR) preionization technique on the newly constructed EXL-50 spherical tokamak, and evaluating the effectiveness on improving the plasma current startup. Two types ECR microwave preionization experiments for the plasma initialization without the central solenoid are reported: (1) 2.45 GHz microwave preionization and current startup with 2.45 GHz ECR source; (2) 2.45 GHz microwave preionization and current startup with 28 GHz ECR source. Application of the 2.45 GHz ECR microwave preionization to the experiments has contributed to (1) getting rid of the plasma breakdown delay; (2) the significant improvement of the discharge quality: the discharge is much longer and more stable while the driven plasma current is larger, compared to the discharge without preionization.