用注入脉冲中性束测量托卡马克中的径向电场

本文分析了用脉冲中性束垂直注入等离子体来探测托卡马克的径向电场分布的方法。该方法是基于对如此形成的离子群体的环向进动速度的测量。磁力线的曲率和电场引起环向漂移。为了把这两种效应分离开来，研究了具有两种能量份额的束的使用。考虑了在实际托卡马克中这种诊断方案的应用。     

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.     

全超导非圆截面托卡马克中性束注入窗口材料活化研究

中性束注入加热为全超导非圆截面托卡马克(EAST)主要辅助加热方式之一.伴随着中性束注入加热,等离子体中子出射强度可达到1014 n/s.由于中性束注入窗口具有较大的开口尺寸,窗口泄露的大量中子可能影响系统的安全稳定运行.本文基于EAST中性束二维模型和蒙特卡洛程序MCNP与材料活化程序FISPACT,研究EAST两条...     

等离子体参数对中性束注入下快离子自举电流影响

本文数值研究了等离子体参数对大纵横比托卡马克装置中性束注入时产生的快离子净电流密度分布的影响。研究表明净电流密度的大小随背景等离子体温度的增大而增大,随等离子体密度的增大而减小,等离子体有效电荷对净电流密度的大小影响较小,但随有效电荷的增大净电流密度的峰值向等离子体边缘偏移。     

HT-7托卡马克实验装置上基于PLC控制的中性束诊断电源系统

文章介绍HT-7托卡马克实验装置上的中性束诊断电源系统的组成、结构，以及基于S7-300PLC的逻辑控制系统。实验表明，该系统操作简单，性能稳定可靠。     

在高密度等离子体内高能中性氢束的衰减

利用对于包括含有激发态的多级磁撞过程的原子过程目前通用的被推荐的截面计算了注入进聚变等离子体的中性氢束的阻止截面和透射。所得到的透射与ＪＴ－６０Ｕ最近的实验很一致。目前的计算表明,多极过程对于在高密度等离本中高能中性氢束的阻止起了关键作用。也给出了对于具有核电荷Ｚ≤８和Ｚ＝２６的等离子体杂质的阻止截面的解析拟合和拟合参数。     

Design of neutral beam injection system for KSTAR tokamak

The neutral beam injection (NBI-1) system has been designed for providing a 300 s deuterium beam of 120 kV/65 A as an auxiliary heating and current drive system of the KSTAR (Korea Superconducting Tokamak Advanced Research) tokamak. The deuterium beam is produced from a long pulse ion source composed of a bucket-type plasma generator and a multi-aperture tetrode accelerator with the help of discharge power supplies and high voltage (HV) power supplies. The beamline components (BLCs) include a neutralizer with an optical multi-channel analyzer (OMA) section, a bending magnet (BM), an ion dump assembly, a movable calorimeter, beam scrapers, and a cryo-sorption pump system in a rectangular vacuum tank. A beam duct equipped with bellows and a voltage break is placed between the NBI vacuum tank and the KSTAR vacuum vessel. All data and parameters of the NBI system are controlled by a control and data acquisition (CODAQ) system through the EPICS based Ethernet interface.     

A scoping study of the application of neutral beam heating on the TCV tokamak

The TCV tokamak contributes to the physics understanding of fusion plasmas, broadening the parameter range of reactor relevant regimes, by investigations based on an extensive use of the existing main experimental tools: flexible shaping and high power real time-controllable electron cyclotron heating (ECH) and current drive (ECCD) systems. A proposed implementation of direct ion heating on the TCV by the installation of a 20–35 keV neutral beam injection (NBI) with a total power of 1–3 MW would permit an extension of the accessible range of ion to electron temperatures (T_i/T_e ～ 0.1–0.8) to well beyond unity, depending on the NBI/ECH mix and the plasma density. A NBI system would provide TCV with a tool for plasma study at reactor relevant T_i/T_e ratios ～1 and in investigating fast ion and MHD physics together with the effects of plasma rotation and high plasma β scenarios. The feasibility studies for a NBI heating on TCV presented in this paper were undertaken to construct a specification for the neutral beam injectors together with an experimental geometry for possible operational scenarios.     

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%.     

An Indian test facility to characterise diagnostic neutral beam for ITER

The diagnostic neutral beam (DNB) line shall be used to diagnose the He ash content in the D–T phase of the ITER machine using the charge exchange recombination spectroscopy (CXRS). Implementation of a successful DNB at ITER requires several challenges related to the production, neutralization and transport of the neutral beam over path lengths of 20.665 m, to be overcome. The delivery is aided if the above effects are tested prior to onsite commissioning. As DNB is a procurement package for INDIA, an ITER approved Indian test facility, INTF, is under construction at Institute for Plasma Research (IPR), India and is envisaged to be operational in 2015. The timeline for this facility is synchronized with the RADI, ELISE (IPP, Garching), SPIDER (RFX, Padova) in a manner that best utilization of configurational inputs available from them are incorporated in the design. This paper describes the facility in detail and discusses the experiments planned to optimise the beam transmission and testing of the beam line components using various diagnostics.     

Argon injection in FT tokamak plasma

H-like argon line spectra have been recorded by means of a high-resolution crystal spectrometer, following puff injection during a tokamak discharge, in a small range of temperature and density plasma conditions (T_e ? 1.2 keV, n_e ? 10¹⁴ cm^?3).Accurate wavelengths determination shows good agreement with theoretical predictions. Second order diffraction of He-like iron, occurring at the same Bragg angle, permits an interesting wavelength comparison.Argon line intensities have been computed, in order to determine both the $\text{Lyα}{}_2\text{Lyα}{}_1$ and the dielectronic satellite to resonance line ratios, the former showing a small increase over the usual 0.5 value towards the end of the discharge and the latter allowing the calculation of a plasma electron temperature in fairly good agreement with Thomson scattering measurements. Finally, the plasma ion temperature is deduced by Doppler broadening of the resonance lines.     

Optical design of the Lyman alpha based beam emission spectroscopy(LAB)diagnostic on the HL-2A tokamak

In this article, we present the optical design of a novel diagnostic on the HL-2 A tokamak, i.e. the20-channel edge Lyman-alpha beam emission spectroscopy, which is a promising solution for edge density turbulence research on tokamaks, as it offers the possibility of density fluctuation measurement with a 3.3 mm spatial resolution while maintains a high temporal resolution of1 μs. The optical path, including the reflective collection optics, the high-dispersion spectrometer, and the linear detector array, is carefully optimized to obtain a good image quality and a high throughput. The maximum root mean square radius of the collection optics is 64 μm.The detected photon flux is estimated to be about 10~(11) photons/s/channel.     

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.     

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.     

蒙特卡罗方法计算介质对窄束γ射线的质量衰减系数

根据γ光子与物质的作用机理，对γ光子在介质中反应轨迹进行抽样跟踪，以此建立基于蒙特卡罗方法的粒子输运数学模型。利用虚拟仪器开发工具LabWindows／CVI编制软件模拟γ光子在介质中的输运过程，根据粒子出射率及对应的吸收介质厚度，计算介质对窄束γ射线的质量衰减系数。结果显示用蒙特卡罗方法计算介质对窄束γ射线的质量衰减系数是可行的。     

Neutral beam deposition in large,finite-beta noncircular tokamak plasmas

A parametric pencil beam model is introduced for describing the attenuation of an energetic neutral beam moving through a tokamak plasma. The nonnegligible effects of a finite beam cross-section and noncircular shifted plasma cross-sections are accounted for in a simple way by using a smoothing algorithm dependent linearly on beam radius and by including information on the plasma flux surface geometry explicitly. The model is bench-marked against more complete and more time-consuming two-dimensional Monte Carlo calculations for the case of a large D-shaped tokamak plasma with minor radiusa=120 cm and elongationb/a=1.6. Deposition profiles are compared for deuterium beam energies of 120–150 keV, central plasma densities of 8×10¹³ to 2×10¹⁴ cm^–3, and beam orientation ranging from perpendicular to tangential to the inside wall.