Engineering design for the HL-2M tokamak components

The HL-2A tokamak will be modified into HL-2M. The Bt at the plasma center (major radius R = 1.78 m) is 2.2 T, the minor radius is 0.65 m. The plasma current I_P of HL-2M will reach up to 2.5 MA, the elongation and triangularity is more than 1.8 and more than 0.5, respectively. The vacuum vessel torus consists of 20 sectors with “D” shaped cross-section and double wall structure. 20 toroidal field coil bundles comprise 140 turns which are designed with demountable joints, the poloidal field coils system consists of 25 coils. The engineering design and calculation for field coil system, vacuum vessel, support structure, etc. are finished, many key issues for manufacture process have been discussed with industry and the fabrication of main components of HL-2M tokamak will be carried out in factories.     

中国环流器二号A(HL-2A)装置物理设计

文章是关于中国环流器二号A(HL-2A)装置物理设计的总结报告,包括以下几方面的内容:分析计算等离子体截面变形及由截面拉长引起的垂直不稳定性,提出对HL-2A极向磁场线圈电流和控制系统的要求;研究通过中性束注入加热(NBI)和低混杂波电流驱动(LHCD)实现等离子体剖面控制,模拟并设计HL-2A的高性能的运行模式;分析HL-2A先进约束位形(RS位形)下的磁流体力学不稳定性,为实现高性能模式稳态运行的等离子体控制指出方向;同时,利用数值模拟分析HL-2A偏滤器等离子体性能,为偏滤器的改进提供依据。     

identification and Analysis of Magnetic Structures on HL-2A

A method for the identification and analysis of magnetic islands is presented based on the calculation of the perturbative current and magnetic field in plasmas. A cylindrical approximation is adopted and the toroidal effect on plasma equilibrium is also included. This method has been used on the HL-2A tokamak for analysing the magnetic island structures.     

HL-2A托卡马克上等离子体环电流、环电压测量系统研制

详细描述了HL-2A托卡马克装置上等离子体环电流、环电压诊断系统,其中包括测量原理、诊断安排、实验等方面的内容。对实验中出现的杂散场干扰问题,及补偿技术做了全面的研究和探讨。并结合HL-2A的实验数据对诊断系统进行了验证。     

HL-2A Tokamak Edge Modeling with B2

The outer divertor plasma of HL-2A and its associated scrape-off plasma have been simulated using a two-dimensional multi-species fluid code of Braams with a simplified neutral gas model. HL-2A has a double-null closed divertor in separate divertor chambers above and below the nearly circular plasma tours. The computed numerical grid is developed according to an ideal magnetic surface. The calculation is involved only with pure hydrogen plasma. The emphasis has been focused on parametric studies involving variation of the assumptions made for the core plasma. The peak temperatures and the heat flux near the target are of the principal concern。     

Imaging System and Plasma Imaging on HL-2A Tokamak

As a new diagnostic means, plasma-imaging system has been developed on the HL-2A tokamak, with a basic understanding of plasma discharge scenario of the entire torus, checking the plasma position and the clearance between the plasma and the first wall during discharge. The plasma imaging system consists of (1) color video camera, (2) observation window and turn mirror, (3) viewing & collecting optics, (4) video cable, (5) Video capture card as well as PC. This paper mainly describes the experimental arrangement, plasma imaging system and detailed part in the system, along with the experimental results. Real-time monitoring of plasma discharge process, particularly distinguishing limitor and divertor configuration, the imaging system has become key diagnostic means and laid the foundation for further physical experiment on the HL-2A tokamak.     

ECRH Launcher for Four-Beam Injection on HL-2A Tokamak

An Electron Cyclotron Resonance Heating (ECRH) launcher for four-beam injection has been installed on the HL-2A tokamak and used in plasma heating experiments. By the launcher, four EC wave beams can be injected into the tokamak through a φ350 mm port, which are generated from 2 sets of 68 GHz/1 s and 2 sets of 68 GHz/1.5 s gyrotron tubes manufactured by GYCOM, with maximum output power of 500 kW for each. In this paper, the properties of the EC beam in the launcher and plasma are presented: at the centre of the cross section of the tokamak, the beam radius is 31.7 mm; thermal analysis of a 3D model indicates that the peak temperature increase would be only 30 o C at the mirror surface for a 500 kW/1 s pulse; ray-tracing calculation predicts satisfactory power deposition. In the plasma experiment, six beams including four beams from this launcher and two from another launcher have been injected simultaneously. Besides, obtaining ELM-y H-mode discharges, the ECRH system shows reliability and stability of the launcher functions.     

Study of Plasma MHD Equilibrium in HL-2A Tokamak

EFIT（Equilibrium Fitting） code is successfully transplanted for HL-2A tokamak parameters. The evolution of the equilibrium configurations for shot 2898 is simulated. It is shown that the discharge starts with a limiter configuration and develops into a divertor configuration gradually after t = 200 ms. The latter lasts until the end of the discharge at t = 900 ms. The evolution of the major plasma shape parameters such as the boundary magnetic fluxes, the positions of the x-point and magnetic axis, and the minor radii obtained are calculated and compared with the experimental results. The agreement between the simulation and experiments are shown to be reasonable. The possibility for discharge quality improvement is discussed.     

The Principle of Plasma Boundary Identification on HL-2A Tokamak

Plasma boundary identification is a basic task for studies on equilibrium and confinement in a divertor tokamak. With the progress on the experiments after engineering experiments, the boundary identification becomes an important issue for HL-2A. In order to satisfy the requirements of preciseness, simplified measurements and quickness, the filament current method instead of solving the equilibrium equations is used to identify plasma boundary on HL-2A. The involved principle, mathematics and the progresses, which have been made with this method, are given.     

The Power Supply System for HL-2A Tokamak＇s Magnetic Field

The highpower pulsed power supply system for the magnetic field of the HL-2A Tokamak is described in this paper. The total output power of its eight magnetic field power supply units of nearly 250 MW. Their highest DC output voltage and current are 3510 V and 45 kA, respectively. All the units are operated in a pulsed mode. The pulse duration is 5 s, and the cyclic period is 15 min. The power supply system consists mainly of pulsed flywheel motor generators, rectifying transformers, thyristor converters, diode rectifiers and switches. The system incorporates many key technologies-supply equalization with two generators and four diode bridges, constant-angle phase triggers with a wide frequency range, current equalization, a status detector for the high current 6-phase converter, and advanced monitoring based on a programmable logic computer and engineering parameter measurement. The experimental results show that the performance of the power supply system satisfies the requirements of HL-2A experiments very well.     

HL-2A托卡马克装置的工程和实验概况

中国环流器二号A装置(HL-2A)是核工业西南物理研究院2002年投入实验运行的托卡马克,它是我国第1个具有偏滤器、等离子体截面具有一定垂直拉长的托卡马克.HL-2A的磁体使用铜导体,具有良好的灵活性和等离子体的可近性,其极向场线圈全部位于环向场线圈之内,位于真空室内的偏滤器的成形线圈可建立双零和单零的偏滤器位形.HL-2A已发展了30多套先进的等离子体诊断系统和总功率4 MW的辅助加热系统,加料技术得到持续发展.随着上述系统的建设和放电综合控制技术的提高,HL-2A装置已获得了高约束模式,这为开展先进托卡马克(AT)物理实验,ITER和聚变堆的科学、技术和工程问题等的研究奠定了基础.HL-2A也成为国际上最活跃的中型托卡马克,为国际托卡马克物理活动(ITPA)作出了积极贡献.     

Fuelling and Diagnostics with Pellet Injection in the HL-1M Tokamak

The pellet injection experiments for fuelling and diagnostics have been carried out on the HL-1M tokamak. The eight-pellet injector was installed on HL-1M. A reliable monitordetector and camera system was set up to take initial pellet photographs and measure the initial pellet speed and size. High fuelling efficiency of 60 % - 100 % and a density profile with a peaking factor of 1.8 - 2.0 were obtained. The maximum density close to 10^14/cm^3 in HL-1M was achieved with newly optimized combined fuelling techniques. Two typical models of pellet ablation have been utilized for simulative calculation of the ablation rates in HL-1M. In comparison with the distribution of the measured Hα emission intensity from the digital data of the CCD camera, the experimental result seems more optimistic for core fuelling than theoretical predictions by the two models. The safety factor profile q（r） has been extracted from the information provided by the CCD camera during the pellet injection. The reliability of the measured results depends mainly on the calibration of the imaging space position. Based on the calibration, the measured q-profile becomes more reasonable than those published previously for the same shot number and same photograph.     

Design of the ECRH/ECCD Launcher System for HL-2A Tokamak

An ECRH/ECCD system with two 68GHz/500kW/1S gyrotrons will be built up in HL-2A tokamak. The location of the Gaussian beam waist is 580 mm away from the center of the plasmas and the beam radius is 37 mm at the center of the plasmas. Compared to the minor plasma radius （420 mm）, it is small enough for localized control. The launcher system covers a wide toroidal and poloidal steering range by the two steering plane mirrors. Therefore it is possible to explore the on- and off-axis heating over half of the plasma minor cross section and the co-current drive.     

The Designed Operation of the Machine Control System on HL-2A Tokamak

The Ethernet and field-bus communications are used in the machine control system （MCS） of HL-2A. The control net, with a programmable logic controller （PLC） as its logic control master, an engineering control management station as its net server, and a timing control PC connected to a number of terminals, flexibly and freely transfers information among the nodes on it with the Ethernet transmission techniques. The PLC masters the field bus, which carries small pieces of information between PLC and the field sites reliably and quickly. The control net is connected into the data net, where Internet access and sharing of more experimental data are enabled. The communication in the MCS guarantees the digitalization, automation and centralization. Also provided are a satisfactory degree of safety, reliability, stability, expandability and flexibility for maintenance.     

Application of DEGAS for Ion Temperature Profile Reconstruction from a NPA Diagnostic on HL-2A

On HL-2A, the ion temperature profile can be roughly estimated by analyzing the energy spectrum of the neutral particles that escape from core plasma through 11 different spatial channels. However, this profile deviates from the real radial distribution because of the averaging and attenuation effects. The DEGAS code is used to estimate these effects and reconstruct a more reasonable ion temperature profile by simulating these neutral spectra and making them conform to the measurements. In this paper, critical parameters that affect the escaped neutral spectra are enumerated, and an example of a reconstruction of the ion temperature profile is presented based on experimental data.     

中国环流器二号A装置工程与实验进展

介绍了中国环流器二号A(简称HL-2A)装置上关键聚变工程技术取得的新进展及在磁约束聚变科学若干关键课题研究取得的重要实验成果,特别是在HL-2A装置上成功实现了偏滤器位形下具有边缘局域模的高约束模式运行,是我国磁约束聚变实验研究史上具有里程碑意义的重大进展,使我国的磁约束聚变科学和等离子体物理实验研究进入了一个崭新的阶段。     

中国环流器二号A装置工程与实验进展

介绍了中国环流器二号A(简称HL-2A)装置上关键聚变工程技术取得的新进展及在磁约束聚变科学若干关键课题研究取得的重要实验成果,特别是在HL-2A装置上成功实现了偏滤器位形下具有边缘局域模的高约束模式运行,是我国磁约束聚变实验研究史上具有里程碑意义的重大进展,使我国的磁约束聚变科学和等离子体物理实验研究进入了一个崭新的阶段。     

Technical Design of Arc-Discharge and Deceleration Power Supply for MW Level NBI System on HL-2A Tokamak

Neutral beam injection (NBI) is one of the most effective ways to heat and drive plasma in a tokamak. The mega watt level neutral beam injector on the HL-2A tokamak consists of four high-power ion sources. Each source is supplied by discharge, beam extraction and auxiliary power supplies. Some circuit topologies and control sequences designed for the system are presented in this paper. Some important technologies such as the notching circuit, insulated gate bipolar transistor (IGBT) series-connected switch, high-frequency switching power supply and control system based on a digital signal processor (DSP) have been used. The system can be effectively used for high current ion beam extraction, protection, ion optics and so on. The power system has been safely used in HL-2A high-parameter NBI experiments for three years. The power of NBI can be kept at higher than 0.75 MW for 1 second and the ion beam power extracted from the ion source is higher than 2 MW. The ion temperature of the plasma center is close to 2.0 keV. These results show that the design of this power system is reasonable and reliable, and it can fully meet the system requirements for NBI of the HL-2A tokamak.