ASDEX中性束注入器的供电系统

本文介绍ＡＳＤＥＸ中性束注入器供电系统概况，对注入器的高压电源，磁场反馈控制和弧打坑电路作较较详细的介绍。     

EAST中性束注入器大厅内辐射剂量计算与测量

利用解析方法对EAST大功率中性束注入器充氢运行时实验大厅内6个关键点的辐射剂量进行了理论计算,并利用光致光剂量计(OSL)对这些位置点进行了辐射剂量测量。理论计算和OSL测量结果表明:理论计算结果与实验测量结果具有一定吻合度。同时还表明:EAST中性束注入器现有的防护装置满足实验运行时辐射防护要求。     

HL-1M中性束注入器等离子体发生器电源

HL-1M中性束注入器和供电方框图如图1所示。注入器主要由离子源、束线和电源等部分组成,而离子源由等离子体发生器(放电室)和加减速结构(引出系统)两部分组成。在这种源型中,等离子体发生器电源包括灯丝、弧流和送气电源,也称做放电电源。     

中性束注入器水流热量计系统优化分析

介绍了用于计算中性束注入实验中束流功率沉积的水流热量计系统优化及优化后的测试结果。前期的水流热量计系统利用串口传输和分散式牛顿模块采集,采样率低、传输速度慢、抗干扰能力差、无法满足实验要求;优化后的系统基于虚拟仪器,采用TCP\IP协议传输和虚拟仪器技术,提高了系统的采样率和精度、优化了数据传输速度及抗干扰能力。优化后的系统经测试可以准确监测中性束注入器装置上各热承载部件冷却水的温升及流量,并分析得到中性束注入时束流在各热承载部件上的功率沉积。实验结果表明优化后的系统工作稳定,使用灵活,数据准确,满足实验要求。     

EAST中性束注入射频离子源放电模拟研究

中性束注入加热是核聚变中非常重要的一种辅助加热手段,离子源所能达到的性能决定了东方超环(Experimental Advanced Superconducting Tokamak, EAST)中性束注入所能达到的指标。为了实现长脉冲和高功率加热的需求,采用射频离子源取代传统的热阴极离子源已成为未来离子源发展的一种趋势。本文对射频离子源的结构设计和放电特性进行了理论模拟研究,给出了线圈匝数、匝间距、驱动器尺寸、放电气压和射频功率等参数与等离子体参数间的关系,为接下来射频离子源的研制和实验奠定一定的理论基础。     

用于ITER中性束注入器的大体积负离子源的改进

对负离子源进行了改进以实现用于国际热核实验堆（ITER）的大功率中性束注入器，已在植入铯的多会切等离子体发生器中在0．1Pa的非常低的气压下产生了31mA.cm^2-(H^-)的强负离子电流密度，该等离子体发生器与ITER源有着相同的概念，对于一个真空绝缘的加速器，已完成了达1．8米的长距离真空间隙的耐压实验，已表明从真空击穿到气体放电的气压距离乘积（pd)的过渡区约为0．2Pa.m，这离ITER源的运行区足够高了，在试验和实验的基础上建造了真空绝缘加速器样机，成功地证实了高能H^-束加速达到970keV,37mA,1s。     

EAST中性束注入器稳态偏转磁铁的参数设计与估算研究

中性束注入器偏转磁铁是剥离束流中剩余离子的关键设备,它与剩余离子吞食器等内部部件构成了中性束注入器的束偏转系统。束偏转系统的性能对中性束注入器束流的品质及其束传输效率发挥着重要作用。本文根据EAST(Experimental Advanced Superconducting Tokamak,EAST)中性束注入器对束偏转系统的要求,对其偏转磁铁各性能参数进行了估算。为中性束注入器设计了一台用以剩余离子180°偏转的偏转磁铁。该偏转磁铁采用H型二极电磁铁结构;其磁极端面设计为138cm×47cm的圆角矩形结构;其线圈设计为每侧2饼,每饼2层,每层8根的串联结构,导线选用外方内圆空心铜导体,以满足偏转磁铁稳态运行的需要。该设计的偏转磁铁在370 A励磁电流条件下,可提供80keV氘离子束偏转所需的磁场。实验测试结果显示:500 A励磁电流稳态运行条件下,偏转磁铁线圈冷却水温升约21.5℃,该设计结构的偏转磁铁满足EAST中性束注入器满参数稳态运行和未来运行参数逐步提高的需要。     

Gas flow and related beam losses in the ITER neutral beam injector

The gas flow in the ITER neutral beam injectors has been studied using a 3D Monte Carlo code to define a number of key parameters affecting the design and operation of the injector. This paper presents the results of calculations of the gas density in the two accelerator concepts presently considered as options for the ITER injectors, and the resultant stripping losses of the negative ions during their acceleration to 1 MeV. The sensitivity of the model to various parameters has been studied, including the gas temperature in the ion source and the subsequent accommodation by collisions with the accelerator structure, and the degree of dissociation of the D₂ or H₂ in the ion source, and subsequent recombination during collisions with the accelerator structure. Additionally the sensitivity of the losses to details of the beam source design and operating parameters are examined for both accelerator concepts.     

EAST-NBI离子源电极打火分析及处理

强流离子源是EAST(Experimental Advanced Superconducting Tokamak)中性束注入器(Neutral Beam Injector,NBI)最关键的核心部件,其能达到的性能在很大程度上决定了EAST中性束注入器所能达到的指标。离子源在束引出时电极打火现象偶有发生,这对于离子源的正常运行有非常严重的影响,甚至危害离子源的寿命。本文结合离子源运行过程中的束引出实验波形和水流量热计(Water Flow Calorimetry,WFC)系统的测量数据得出等离子体发射面的束流光学系统一直处于非最佳聚焦状态是导致打火的原因,试通过优化高压投入时刻等离子体与高压的匹配,实现高压的稳定投入有效抑制打火现象的发生,并且给离子源加入硬件保护机制,为离子源安全稳定运行奠定基础。     

Criticality in the fabrication of ion extraction system for SST-1 neutral beam injector

For the heating of plasma in steady-state superconducting tokamak (SST-1) (Y.C. Saxena, SST-1 Team, Present status of the SST-1 project, Nucl. Fusion 40 (2000) 1069–1082; D. Bora, SST-1 Team, Test results on systems developed for the SST-1 tokamak, Nucl. Fusion 43 (2003) 1748–1758), a neutral beam injector is provided to raise the ion temperature to 1 keV. This injector has a capability of injecting hydrogen beam with the power of 0.5 MW at 30 keV. For the upgrade of SST-1, power of 1.7 MW at 55 KeV is required. Further, beam power is to be provided for a pulse length of 1000S. We have designed a neutral beam injector (S.K. Mattoo, A.K. Chakraborty, U.K. Baruah, P.K. Jayakumar, M. Bandyopadhyay, N. Bisai, Ch. Chakrapani, M.R. Jana, R. Onali, V. Prahlad, P.J. Patel, G.B. Patel, B. Prajapati, N.V.M. Rao, S. Rambabu, C. Rotti, S.K. Sharma, S. Shah, V. Sharma, M.J. Singh, Engineering design of the steady-state neutral beam injector for SST-1, Fusion Eng. Des. 56 (2001) 685–691; A.K. Chakraborty, N. Bisai, M.R. Jana, P.K. Jayakumar, U.K. Baruah, P.J. Patel, K. Rajasekar, S.K. Mattoo, Neutral beam injector for steady-state superconducting tokamak, Fusion Technol. (1996) 657–660; P.K. Jayakumar, M.R. Jana, N. Bisai, M. Bajpai, N.P. Singh, U.K. Baruah, A.K. Chakraborty, M. Bandyopadhyay, C. Chrakrapani, D. Patel, G.B. Patel, P. Patel, V. Prahlad, N.V.M. Rao, C. Rotti, V. Sreedhar, S.K. Mattoo, Engineering issues of a 1000S neutral beam ion source, Fusion Technol. 1 (1998) 419–422) satisfying the requirements for both SST-1 and its upgrade. Since intense power is to be transported to SST-1 situated at a distance of several meters from the ion source, the optical quality of the beam becomes a primary concern. This in turn, is determined by the uniformity of the ion source plasma and the extractor geometry. To obtain the desired optical quality of the beam, stringent tolerances are to be met during the fabrication of ion extractor system.

SST-1 neutral beam injector is based on positive ion source. The extraction system consists of three grids, each having extraction area of (width) 230 mm × (height) 480 mm and 774-shaped apertures of 8-mm diameter. To obtain horizontal focal length of 5.4 m and vertical of 7 m, each grid consists of two halves with 387 apertures. Two halves are inclined at an angle of 1.07 ± 0.01°. For long pulse operation, active water cooling is provided by in-laid down of dense network of 22 wavy semicircular (r = 1.1 ± 0.05 mm) cooling channels in the space available between the apertures. The required flatness of the copper plate is 100 μm and positioning tolerance of aperture is ±60 μm. The measurement obtained after fabrication is compared with the specifications. It is pointed out that fabrication within set tolerance limit could be achieved only through process of fabrication and high-resolution measurements.     

Thermal analysis of EAST neutral beam injectors for long-pulse beam operation

Two sets of neutral beam injectors(NBI-1 and NBI-2) have been mounted on the EAST tokamak since 2014. NBI-1 and NBI-2 are co-direction and counter-direction, respectively. As with indepth physics and engineering study of EAST, the ability of long pulse beam injection should be required in the NBI system. For NBIs, the most important and difficult thing that should be overcome is heat removal capacity of heat loaded components for long-pulse beam extraction. In this article, the thermal state of the components of EAST NBI is investigated using water flow calorimetry and thermocouple temperatures. Results show that(1) operation parameters have an obvious influence on the heat deposited on the inner components of the beamline,(2) a suitable operation parameter can decrease the heat loading effectively and obtain longer beam pulse length, and(3) under the cooling water pressure of 0.25 MPa, the predicted maximum beam pulse length will be up to 260 s with 50 keV beam energy by a duty factor of 0.5. The results present that, in this regard, the EAST NBI-1 system has the ability of long-pulse beam injection.     

Design of the reflection magnet and its shielding effect analysis for the neutral beam injector of EAST

In this paper,a reflection magnet to be installed in the EAST neutral beam injection system is simulated and designed.The field intensity of reflection magnet of 42-cm maximum bending radius is about 1.539×10-1 T for 100 keV deuterium beam.The shielding cage is formed by rods.Using the ANSOFT software,the magnetic shielding effect was estimated at about 3% at the magnet pole region.     

Design study of a 120-keV,3He neutral beam injector

We describe a design for a 120-keV, 2.3-MW,³He neutral beam injector for use on a D-³He fusion reactor. The constraint that limits operating life when injecting He is its high sputtering rate. The sputtering is partly controlled by using an extra grid to prevent ion flow from the neutralizer duct to the electron suppressor grid, but a tradeoff between beam current and operating life is still required. Hollow grid wires functioning as mercury heat pipes cool the grid and enable steady state operation. Voltage holding and radiation effects on the acceleration grid structure are discussed. We also briefly describe the vacuum system and analyze use of a direct energy converter to recapture energy from unneutralized ions exiting the neutralizer. Of crucial importance to the technical feasibility of the³He-burning reactor are the injector efficiency and cost; these are 53% and $5.5 million, respectively, when power supplies are included.The beam is composed of 91 separate, parallel currents that flow in the gaps between the elements or wires of a grid. Each such flow is referred to as a beamlet. The current densities in Figs. 5, 8, and 9 are values within a beamlet, as measured at the beam-forming grid. They are not values averaged over the entire beam cross-section.     

EAST托卡马克的中性束注入方案

高能中性束注入(Neutral beam injection,NBI)是核聚变装置托卡马克采用的芯部辅助加热和非感应电流驱动主要手段之一.本文介绍了国家大科学工程全超导托卡马克实验装置(Experimental advanced super-conductingtokamak,EAST)上的高能NBI加热方案及注入器的工程要求,并讨论了中性束在EAST等离子体中的传输等相关问题.     

Achievement of 1000 s plasma generation of RF source for neutral beam injector

A radio frequency(RF) ion source was developed for neutral beam injector in Institute of Plasma Physics, Chinese Academy of Sciences(ASIPP). A cylindrical driver based RF plasma source was tested and optimized for long pulse operation. Recently, the plasma source achieved 1000 s stable plasma discharge with RF power of 35 k W and source pressure of 0.7 Pa for the first time.The heat loading on each of driver components such as the Faraday shield(FS), RF coil and expansion chamber was measured by the water flow calorimeter system. The experiment results showed that FS is the component with highest heat loading and the heat loading on FS was about 65.8% of total heat loading on source. The details of the long pulse operation of RF plasma source are presented in this paper.     

Optimization of the gas flow in the neutralization region of EAST neutral beam injector

According to the problems encountered in the experiments of the EAST neutral beam test stand, the design of neutralizer of EAST neutral beam injector is suggested to modify to optimize the gas flow in the neutralization region. The modifications contain narrowing the slits between the neutralizer and the mounting flange hole, and rotating the gas injection angle from 90° to 60° in the neutralizer. In this paper, an adjusted Direct Simulation Monte Carlo (DSMC) code was used to estimate the modification. The results show that a little change of the slits width causes a large variation of gas target thickness, and the rotation of the gas injection angle can effectively reduce the gas density near the accelerator but with a little of decrease of target thickness.