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.     

EAST中性束注入下快离子输运行为的研究

利用等离子体输运分析程序TRANSP和粒子导心轨道模拟程序ORBIT,结合聚变中子和等离子体储能诊断数据,对EAST(Experimental Advanced Superconducting Tokamak)托卡马克中性束注入加热时不同等离子体电流和纵场强度下的快离子输运行为进行了研究。实验和模拟结果表明:中性束反向束注入时的快离子初始轨道损失功率大于中性束同向束;中性束4条束线产生快离子初始轨道损失区域主要集中在装置中平面以下第一壁以及偏滤器区域;在EAST目前的运行参数范围内,提高等离子体电流和纵场强度,可以使等离子体中快离子的漂移轨道宽度和拉莫尔回旋半径缩小,更有利于快离子的约束,从而提高中性束的加热效率,增加聚变中子产额和等离子体储能。     

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。     

在欧洲联合环等离子体中杂质的极向不对称分布

在欧洲联合环最佳化的剪切托卡马克等离子体中，在镍激光注入实验之后，实验上已经观测到镍离子的极向不对称分布。两类不对称性出现，在高约束模等离子体中早已观测到的第一类不对称性中，镍离子积聚在极向截面的外侧，这可由中性束注入驱动的快环向等离子体转动予以充分解释。第二类不对称性在位置上是相反的：在射频加热的最佳化剪切等离子体中，已观测到镍离子积聚在极向截面的内侧上。     

在JT—60U装置上的快粒子实验

本文介绍了在JT－60U装置上的快粒子实验结果。快粒子是通过ICRF加热和基于负离子的中性束注入（N－NBI）加热产生的。两种加热系统激发了阿尔芬本征模（AE）,在锯齿致稳实验中,常用ICRF激发的AE摄取等离子体中心的磁安全分布信息,将这些测量结果与动态斯塔克（谱线磁裂）效应测量结果及锯齿模型作了比较,还给出了TAE径向模结构的首批测量结构,正如用X模反射仪系统测量的结果一样,当把N－NBI束注入到锯齿等离子体中时,观察到锯齿周期大幅增加,研究了注入N－NBI时,在阿尔芬频率范围内发现的猝发模,线性调频 稳态频率模,这些模式中的大多数,而非所有的模式,是阿阿尔芬速度变化的,从连续波开始并线性变化到环向阿尔芬本征模（TAE）间隙（一般为200ms) 的这样一组模可被认定为共振TAE,而另一组模则在阿尔芬连续波中以短脉冲（达20ms)出现。     

EAST托卡马克激光吹气杂质注入系统控制系统设计及测试

开展托卡马克等离子体中杂质输运与杂质控制研究对于提升等离子体约束性能与保障装置安全有重要意义。为了在EAST(Experimental Advanced Superconducting Tokamak)托卡马克装置上开展杂质输运研究,需要发展一套用于注入示踪杂质粒子的激光吹气（Laser Blow-off,LBO）微量杂质注入系统。本文描述了一种用于EAST托卡马克装置上LBO系统的控制系统设计过程与测试结果。该系统采用了全新的自动控制流程,使得系统可以重复、定量地向EAST等离子体注入不同元素的示踪杂质粒子。该设计通过STM32单片机系统实现对聚焦透镜位移、激光器被触发时刻的准确控制,且激光光斑直径可调,以改变杂质注入量。测试结果显示,系统能快速检测到外触发信号并实现精确定时,激光光斑聚焦位置误差小于0.40 mm,达到激光吹气杂质注入实验要求。本研究对在EAST装置上开展等离子体杂质输运研究具有重要意义。     

HL-1M边缘区径向电场对等离子体转动的影响

我们曾经用计算程序ROTATE编码对HL-1M托卡马克常规运行参数下的主离子与杂质离子的环向转动速度和极向转动速度进行理论模拟计算和比较研究。在这个计算程序里我们挑选了3个具有代表性的等离子体转动理论模型:Hazeltine的标准新经典理论模型,Kim等人区分主离子和杂质离子的转动理论模型,Hinton等人考虑电场梯度对离子转动速度的影响,引     

托卡马克边界等离子体数值模拟

用合适的模拟程序,通过数值求解二维多流体等离子体输运方程来模拟托卡马克边界等离子体输运过程和输运特性。模拟结果能显示边界层区域等离子体参数的分布特性,尤其能显示第一壁和偏滤器靶板附近等离子体参量的分布特性。模拟计算的结果对托卡马克     

托卡马克边界等离子体数值模拟(英文)

用合适的模拟程序,通过数值求解二维多流体等离子体输运方程来模拟托卡马克边界等离子体输运过程和输运特性。模拟结果能显示边界层区域等离子体参数的分布特性,尤其能显示第一壁和偏滤器靶板附近等离子体参量的分布特性。模拟计算的结果对托卡马克偏滤器和第一壁的设计有重要意义。     

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。     

Results of Predictive Fokker?CPlanck Modelling of NBI Deuterons in ITER

First an analytical formalism is presented for calculating the source distribution of ions generated by neutral beam injection (NBI) in tokamak plasmas. A general NBI ion source term, applicable to studies in the phase space up to 6 dimensions, is provided for neutral beams with finite thickness and divergence. Further, using this source term for the envisaged NBI in ITER, we carry out 3D Fokker?CPlanck modelling of the steady-state deuteron distribution function of NBI produced fast deuterons relaxing on bulk plasma components. For two basic ITER scenarios we demonstrate the poloidal profiles of the beam deuteron density, of the NBI generated current as well as of the NBI power deposition to bulk electrons and ions. Further, we evaluate the capability of gamma and NPA diagnostics of NBI ions in ITER and demonstrate the sensitivity of the distributions of NBI generated ions to different ITER operation scenarios.     

HL-2AMW级中性束注入系统弧流电源设计

中性束注入是磁约束受控核聚变实验装置中加热等离子体最有效的方法之一。针对中国环流器2号中性束大功率离子源的特点,从系统功能、主电路拓扑结构、控制硬件及控制时序等方面对弧流电源进行设计。整个电源由低位移相交流调压、高压隔离降压变压器、整流滤波和电流快速转移电路4部分组成。IGBT与电阻串联组成电流快速转移阵列电路,与离子源并联,可实现电流单次或多次快速转移、参数远程设定,有效用于强流离子束的引出和保护。选用DSP和CPLD电路技术实现低位与高位控制器。实验数据显示,该电源最大输出为200kW/1000A,纹波小于2％,开关上升下降时间达μs级。目前,该电源已安全运行3年,可靠性高,完全满足装置离子源及系统要求,也可应用于其它等离子体技术应用场合。     

Simulations of first-orbit losses of neutral beam injection(NBI) fast ions on EAST

Simulations of first-orbit losses of neutral beam injection(NBI) fast ions in the EAST tokamak have been studied in detail by using the orbit-following code GYCAVA and the NBI code TGCO. Beam ion losses with the wall boundary are smaller than those with the last closed flux surface boundary. In contrast to heat loads on the wall without radio frequency wave(RFW)antennas, heat loads on the wall with RFW antennas are distributed more locally near the RFW antennas. The direction of the toroidal magnetic field dramatically affects the final positions of lost fast ions, which is related to the magnetic drift. The numerical results on heat loads of beam ions corresponding to different toroidal magnetic fields are qualitatively consistent with the experimental results. Beam ion losses increase with the beam energy for the co-current NBIs and the counter-perpendicular NBI. We have studied the behavior of fast ions produced by a small section neutral beam(beamlet) by using the numerical tool NBIT. The distributions of the loss fraction of beamlet fast ions peaked near the edge of the beam section for the counter-current NBIs, and they are related to the injection angle. This indicates that the first-orbit losses can be reduced by changing the shape of beam cross section.     

Effect of Plasma Rotation on Neutral Beam Heating and Current Drive in Tokamaks

For a rapidly rotating plasma, the effects of the resulting Doppler shift have to be included in the neoclassical theory of neutral beam heating, current drive, and plasma transport. In this paper, an improved simulation of neutral beam injection (NBI) and current drive in rotating plasmas is introduced. NBI is simulated using the Monte Carlo code NUBEAM along with the transport code ONETWO. The physical characteristics of heating and current drive for co- and counter-NBI are investigated for non-rotating, co-rotating, and counter-rotating plasmas, all of which can take place in the experiments. In general, it is found that rotation of the plasma can increase the NBI power deposition on the plasma electrons but has little effect on the ions. Moreover, plasma heating by co-NBI is more efficient than that by counter-NBI. For neutral beam current drive, because of the Doppler shift, co-rotation (counter-rotation) of the bulk plasma tends to decrease the co-NBI (counter-NBI) driven current. On the other hand, due to trapping and orbit loss of the fast ions, co-rotation (counter-rotation) has little effect on the counter-NBI (co-NBI) driven current. The results are applied to the forthcoming NBI heating and current drive experiments of the EAST tokamak and should also be useful in the design of experiments in ITER.     

Commissioning of the first KSTAR neutral beam injection system and beam experiments

The neutral beam injection (NBI) system was designed to provide plasma heating and current drive for high performance and long pulse operation of the Korean Superconducting Tokamak Advanced Research (KSTAR) device using two co-current beam injection systems. Each neutral beam injection system was designed to inject three beams using three ion sources and each ion source has been designed to deliver more than 2.0 MW of deuterium neutral beam power for the 100-keV beam energy. Consequently, the final goal of the KSTAR NBI system aims to inject more than 12 MW of deuterium beam power with the two NBI for the long pulse operation of the KSTAR. As an initial step toward the long pulse (～300 s) KSTAR NBI system development, the first neutral beam injection system equipped with one ion source was constructed for the KSTAR 2010 campaign and successfully commissioned. During the KSTAR 2010 campaign, a MW-deuterium neutral beam was successfully injected to the KSTAR plasma with maximum beam energy of 90 keV and the L-H transition was observed with neutral beam heating. In recent 2011 campaign, the beam power of 1.5 MW is injected with the beam energy of 95 keV. With the beam injection, the ion and electron temperatures increased significantly, and increase of the toroidal rotation speed of the plasma was observed as well. This paper describes the design, construction, commissioning results of the first NBI system leading the successful heating experiments carried in the KSTAR 2010 and 2011 campaign and the trial of 300-s long pulse beam extraction.     

Simulations of NBI fast ion loss in the presence of toroidal field ripple on EAST

NBI fast ion losses in the presence of the toroidal field ripple on EAST have been investigated by using the orbit code GYCAVA and the NBI code TGCO. The ripple effect was included in the upgraded version of the GYCAVA code. It is found that loss regions of NBI fast ions are mainly on the low field side near the edge in the presence of ripple. For co-current NBIs, the synergy effect of ripple and Coulomb collision on fast ion losses is dominant, and fast trapped ions located on the low field side are easily lost. The ripple well loss and the ripple stochastic loss of fast ions have been identified from the heat loads of co-current NBI fast ions. The ripple stochastic loss and the collisioninduced loss are much larger than the ripple well loss. Heat loads of lost fast ions are mainly localized on the right side of the radio frequency wave antennas from the inside view toward the first wall. For counter-current NBIs, the first orbit loss due to the magnetic drift is the dominant loss channel. In addition, fast ion loss fraction with ripple and collision for each NBI linearly increases with the effective charge number, which is related to the pitch angle scattering effect.     

Numerical study on the loss of fast ions produced by minority ion cyclotron resonance heating in EAST

The classical prompt loss of fast ions produced by minority ion cyclotron resonance heating(ICRH)is studied by a guiding center orbit following code in the Experimental Advanced Superconducting Tokamak(EAST).It is found that the loss of fast ions produced by ICRH mainly appears in both ends of the resonance layer,while the loss of fast ions in the middle resonance layer is very small.The dominant fast loss comes from trapped ions,rather than from passing ions.Controlling the location of resonance layer at the plasma core may be more beneficial to the EAST tokamak ICRH.In addition,the loss distribution of fast ions is studied.The results show that the fast ions are mainly lost near the midplane in the poloidal direction,but almost uniformly in the toroidal direction.Moreover,we investigate the dependence of fast ion loss on the ICRH power.The simulation results show that the loss fraction of fast ions in both ends of the resonance region increases with the ion cyclotron range of frequencies(ICRF)power,but barely affects the loss of fast ions in the middle region.     

EAST强流离子源电源系统的初步测试运行

测试NBI大功率强流离子源的综合测试台正在建设,已研制了离子源等离子体发生器电源系统、等离子体电极高压电源及梯度极分压器、抑制极负高压电源等电源系统,以及高压传输线及缓冲器,在测试台上开展了对EAST中性束注入器第一台兆瓦级强流离子源样机进行整体电源系统测试和离子源起弧放电的初步测试,完成了离子源电源系统初步性能测试及...     

Edge Structure of Reynolds Stress and Poloidal Flow on the HL-1M Tokamak

1. IntroductionThe determination Of electrostatic Reynolds stressand plasma poloidal flow velocity in scrape-off 18yer(SOL) and on the boundary of tokajxnak plasma havebeen of prime importance due to its potential rolein confinement and the L-H mode transition [1-5].As the plasma confinement is sensitive to the edgeconditions, various mechanisms have been theoretically proposed to explain the creation of a shearedpQloidal flow [6-8]. In brief, the theories attemptingto explain the L--H tra…     

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.