大功率NBI系统的PLC时序控制应用

论述了利用PLC逻辑关系对放电实验控制运行的工作原理,介绍了PLC梯形图应用程序和VB6.0环境下上位机监控程序的开发以及良好的人机操作界面,通过上位机监控界面来实时监控各电源及设备的运行状态.PLC控制系统的应用保障了实验装置的安全运行,极大地方便了实验中对放电参数的改变和设置.     

HT-7超导托卡马克上聚变中子产额和中子注量率分布计算

利用所编制的三维托卡马克中子计算程序,计算了HT-7装置上中子产额随等离子体中心离子温度与密度的变化.同时计算了中子注量率及其在空间不同位置与方位角的分布特性.考虑了离子温度与密度分布的形状因子不同所带来的误差.计算结果为中子诊断和中子辐射防护等提供了依据.     

Experimental Study of Diagnosis Neutral Beam for HT-7 Tokamak

Parametrical effect on plasma discharge and beam extraction in the diagnosis neutral beam （DNB） system for HT-7 tokamak was studied experimentally. Useful results with an improved beam quality were obtained.     

Control System of Neutral Beam Injection on HT-7

Neutral Beam Injection control system (NBICS) is constructed to measure the plasma current, Magnet current, vacuum pressure, cryopump temperature, control water cooling, filament voltage, and power supply， etc. The NBICS, consisting mainly of a Programmable Logic Controller (PLC) subsystem, data acquisition and processing subsystem and cryopump and vacuum pressure monitoring subsystem, has successfully been used on a NBI device. In this article, the design of NBICS on HT-7 is discussed and each subsystem is described in particular.In addition, some experimental results are reported which are very important data for further research related to the HT-7 tokamak.     

Logical and Timing Control for Diagnostic Neutral Beam Injection on HT-7

The timing and master control logic （MCL） units are the most important function units of the diagnostic neutral beam （DNB） power supply control system. The units control the operation of nine power supply subsystems of the DNB system, and provide protection for the DNB system from faults such as beam source arc down. Based on the characteristics of the DNB power supply system, the timing and MCL units have been designed, fabricated and tested. Experiments prove that the timing unit is convenient, flexible and reliable, and the MCL is functional.     

Analysis on Pressure Distribution in HT-7 Neutral BeamInjection System

Neutral Beam Injection (NBI) is an effective way to improve the efficiency of tokamak heating system. This article primarily introduces a work on the pressure distribution inside the tank of NBI heating system, especially inside the neutralizer, which is got by selecting a proper mathematical model and constructing a series of rational calculating formulas on pressure distribution. Furthermore, we simulate the pressure distribution by the Monte Carlo method. Comparing the result of simulation with that of theoretical calculation, we find that both the results are very close each other, showing their mutual validity.     

放疗加速器束流剂量实时成像系统

介绍了一种放疗加速器束流剂量实时成像系统,采用数字化脉冲X射线成像技术获取放疗加速器的束流强度分布图像,并利用数字图像处理技术生成实时3D剂量图。利用该系统可以实时、直观地获得加速器束流强度的相对分布曲面,给出任意剖面的束流强度分布曲线和FWHM、半影、峰值、对称度、均整度等参数。利用本系统的实时性,在国内首次实现了放疗加速器动态楔形剂量分布和束流过渡过程的测量。给出了系统配置和实测的放疗加速器束流强度分布图像。     

Upgrade of plasma density feedback control systemin HT-7 tokamak

The HT-7 is a superconducting tokamak in China used to make researches on the controlled nuclear fusion as a national project for the fusion research. The plasma density feedback control subsystem is the one of the subsystems of the distributed control system in HT-7 tokamak (HT7DCS). The main function of the subsystem is to control the plasma density on real-time. For this reason, the real-time capability and good stability are the most significant factors, which will influence the control results. Since the former plasma density feedback control system (FPDFCS) based on Windows operation system could not fulfill such requirements well, a new subsystem has to be developed. The paper describes the upgrade of the plasma density feedback control system (UPDFCS), based on the dual operation system (Windows and Linux), in detail.     

Analysis of charge-exchanged neutral particles during IBW heating in the HT-7 tokamak

The charge-exchange neutral particles fluxes and energy distribution in IBW heated plasma were investigated in the HT-7 tokamak. The RF frequency was 30 MHz and with an injecting power up to 200 kW. It is observed that the plasma performance is obviously enhanced by IBW heating. The electron temperature was increased by 0.5 keV and the central line averaged electron density was doubled. The neutral particle fluxes of high-energy increased and the bulk ions were heated during IBW heating. The ion temperature was increased by 0.3 keV and the ion heating efficiency of (2–3) eV kW⁻¹ × 10¹³ cm⁻³ was achieved. The velocity distribution of charge-exchanged neutral particles appears to be Maxwellian without high-energy tail ions up to the maximum RF power.     

On the comparison of three methods of assessing beam quality for broad beam in vitro cell irradiation

The interaction of charged particles with living matter has recently attracted increasing interest in the field of biomedical applications such as hadron therapy, radioprotection and space radiation biology. Particle accelerators are particularly useful in this area.In vitro radiobiological studies with a broad beam configuration require beam homogeneity. The goal is to produce a dose distribution given to a cell population that is as close to uniform as possible.In this paper, we compare the results of three devices used to assess the beam quality for broad beam irradiation: a passivated implanted planar silicon (PIPS) particle detector, a position-sensitive solid state detector, which is camera-like, and a solid state nuclear track detector (CR39).The first device is a PIPS detector of 300 μm nominal depletion depth and an entrance window with a thickness of about 500 Å. It is collimated with a 0.5 mm aperture and mounted in air on an XY moving table as close as possible to the exit window of the beam line.The second device is a CMOS position-sensitive detector (technological process 0.6 μm AMS CUA), 112 × 112 pixels, with 153 × 153 μm² pixel size. It allows the user to rapidly obtain dose uniformity over a surface of 1 × 1 cm². During uniformity and dose rate assessment it is placed in air at the PIPS location.For both detectors, beam profile was obtained for various proton fluxes (from ∼5 × 10⁴ to 10⁶ particles cm⁻² s⁻¹). Preliminary tests were made with CR39 using 4 MeV He⁺⁺ ions.Results are analysed using Poisson distribution and cell hit probability.     

Design of flowing liquid lithium device for HT-7 tokamak

Lithium is a very attractive element due to its very low radiation power, strong H retention as well as strong O getter activity. Flowing liquid lithium (FLiLi) device, to be used as a plasma-facing limiters, has been designed and will be tested in HT-7 tokamak. It is mainly composed of distributor, guide plate, collector, and heater as well as cooling loop. The heater uses heater strip and cooling loop design, to control the temperature of lithium on the guide plate ranging from 200 °C to 400 °C. The distributor attached to feeding pipe, distributes liquid lithium (LiLi) flowing on the guide plate. The collector was designed to reclaim the superfluous LiLi and transport it out of device.The paper focuses on the design of flowing liquid lithium device. In addition to the process of design, thermal analysis has been carried out using finite element method (FEM) for optimizing the structure of heater and cooling loop and results of analysis are presented.     

NBI双潘宁离子源参数的静电探针测量

介绍了静电探针在22cm双潘宁离子源中心等离子体参数测量中的应用.测出了探针的伏安特性曲线,由该曲线求出在弧流为200 A时,电子温度约为6.0 eV,密度约为2×1012 cm-3,这对于调整放电控制条件,获得高效率大功率的离子源有重要的参考价值,并且对于今后实现强而且均匀的中性束束流的引出也有着重要意义.     

HT-7诊断中性束评价系统温升特性分析

本文介绍了HT-7诊断中性束装置束评价系统的基本结构和工作原理;分析靶板采样孔的遮挡效应、束腰位置和前、后靶上采样铜柱的温升比值的关系,解释了相同参数下后靶板温升较高的实验现象;根据前、后靶板的平均温升比值,计算了装置运行中的束流焦距,结果显示其误差不大于5%。     

诊断中性束电源的数据采集与处理系统

文中介绍了一种用于诊断中性束电源系统的数据采集与处理系统(DPAS).在电源系统的运行过程中,系统需要对18路电流/电压信号进行实时采集.DPAS采用网络技术以客户端/服务器模式进行开发.系统引进了一种高效的压缩算法从而大量地节约了数据存储所需的磁盘空间.系统所采用的数据存储格式与EastScope软件(一款用于查看实验数据的软件)兼容,从而使实验操作人员可以很容易对实验结果进行分析和总结.     

剩余气体束流剖面探测系统的初步研制

介绍了剩余气体柬流剖面探测系统的工作原理和设计依据，描述了所设计的系统结构并进行了初步实验，证明了利用剩余气体探测柬流剖面的方法是可行的。     

图像采集及处理技术在粒子加速器束流剖面测量中的应用

介绍了ＨＩＲＦＬ（兰州重离子研究装置）上利用荧光靶对束团截面形状和流强大小的实时测量方法和对光斑图像进行动态分析的硬件结构以及软件的设计方法,并定量地给出一些测量结果。该系统具有可靠性好、精度高、测量直观、应用软件的界面友好等优点。     

Simulation of supersonic molecular beam fueling on HL-1M tokamak

The supersonic molecular beam (SMB) ablation and penetration processes in the HL-1M tokamak experiments are studied. The cluster formation and dissolution, SMB adiabatic expansion, shielding and cooling effect are all taken into account. An optimized numerical modelis applied in the analysis and shown to be in good agreement with the experimental observations. The possibility of fueling large tokamak plasmas with SMB injection is explored.     

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.     

Liquid lithium surface control and its effect on plasma performance in the HT-7 tokamak

Experiments with liquid lithium limiters (LLLs) have been successfully performed in HT-7 since 2009 and the effects of different limiter surface structures on the ejection of Li droplets have been studied and compared. The experiments have demonstrated that strong interaction between the plasma and the liquid surface can cause intense Li efflux in the form of ejected Li droplets – which can, in turn, lead to plasma disruptions. The details of the LLL plasma-facing surface were observed to be extremely important in determining performance. Five different LLLs were evaluated in this work: two types of static free-surface limiters and three types of flowing liquid Li (FLLL) structures. It has been demonstrated that a FLLL with a slowly flowing thin liquid Li film on vertical flow plate which was pre-treated with evaporated Li was much less susceptible to Li droplet ejection than any of the other structures tested in this work. It was further observed that the plasmas run against this type of limiter were reproducibly well-behaved. These results provide technical references for the design of FLLLs in future tokamaks so as to avoid strong Li ejection and to decrease disruptive plasmas.     

Plasma profile and shape optimization for the advanced tokamak power plant, ARIES-AT

An advanced tokamak plasma configuration is developed based on equilibrium, ideal MHD stability, bootstrap current analysis, vertical stability and control, and poloidal field coil analysis. The plasma boundaries used in the analysis are forced to coincide with the 99% flux surface from the free-boundary equilibrium. Using an accurate bootstrap current model and external current drive profiles from ray tracing calculations in combination with optimized pressure profiles, β_N values above 7.0 have been obtained. The minimum current drive requirement is found to lie at a lower β_N of 6.0. The external kink mode is stabilized by a tungsten shell located at 0.33 times the minor radius and a feedback system. Plasma shape optimization has led to an elongation of 2.2 and triangularity of 0.9 at the separatrix. Vertical stability could be achieved by a combination of tungsten shells located at 0.33 times the minor radius and feedback control coils located behind the shield. The poloidal field coils were optimized in location and current, providing a maximum coil current of 8.6 MA. These developments have led to a simultaneous reduction in the power plant major radius and toroidal field from those found in a previous study [S.C. Jardin, C.E. Kessel, C.G. Bathke, D.A. Ehst, T.K. Mau, F. Najmabadi, T.W. Petrie, the ARIES Team, Physics basis for a reversed shear tokamak power plant, Fusion Eng. Design 38 (1997) 27].