Design, fabrication and testing of UHV compatible high power RF devices for lower hybrid current drive system on SST-1 tokamak

Maintaining plasma current under steady state conditions is one of the most important pre-requisites for a tokamak-based reactor. Lower hybrid current drive (LHCD) system aims to drive tokamak plasma current by means of RF power. The LHCD system on SST-1 tokamak is based on two 500 kW, CW klystrons operating at 3.7 GHz. A waveguide transmission line transmits power from source to the antenna. A phased array waveguide antenna is used to couple power to the plasma. The antenna side of the transmission line is placed inside the tokamak vacuum vessel. The design and fabrication of this In-Vessel system has to satisfy the demands of high power RF as well as ultra high vacuum (UHV) compatibility. This paper describes some of the critical UHV compatible In-Vessel RF devices, their design, fabrication, and test results.     

Experimental Investigation of Lower Hybrid Wave Coupling in HT-7

Experiments on lower hybrid wave(LHW)coupling were investigated in the HT-7 tokamak.Good coupling of LHW plasma has been demonstrated at different conditions in the HT-7 tokamak.Relevant results have proved that LHW-plasma coupling is affected by the phase difference between adjacent waveguides.Furthermore,the edge density around the grill and relevant coupling can be adjusted by changing the plasma line average density or the gap value between the LH grill and the last closed flux surfaces(LCFS).It is found that the coupling of LHWs becomes poor when the edge density around the LH grill is large enough in the HT-7tokamak,and that coupling remains good with a proper edge density.With increasing LHW power,it is also found that the reflection coefficients(RCs)increase due to non-linear effects under conditions of low edge recycling,but can decrease under high edge recycling.The edge density depends mainly on the competition between the ponderomotive force(PMF)and the edge recycling intensity in the HT-7 tokamak.     

Analysis of high β regimes for DEMO

Stability limit calculations are presented for a range of tokamak power plant equilibria. The current drive requirements to sustain the optimised equilibrium profiles are confirmed by a transport code and the plasma shape is obtained from free-boundary equilibrium calculations. A pressure pedestal is included according to empirical scaling and ballooning mode stability limits. A terative optimisation of the profiles is undertaken to improve the baseline profiles in order to achieve the highest possible plasma performance and most favourable magnetohydrodynamic stability within conservative assumptions in order to increase confidence in the availability and control of the plasma. This results in a fully noninductive baseline operating scenario for a tokamak power plant design which has a broad low-shear q-profile which is meant to complement previous advanced tokamak design studies.     

ISTTOK plasma control with the tomography diagnostic

The presently available processing power in generic processing units (GPU) combined with state-of-the-art programmable logic devices enables the implementation of complex algorithms for plasma diagnostics in a real-time scenario.A tomography diagnostic based on three linear pin-hole cameras each with eight lines of sight has been developed for the ISTTOK tokamak. The plasma emissivity in a poloidal cross-section is computed locally on a sub-millisecond time scale, using a variant of the Fourier-Bessel algorithm. The output signals are then used for active plasma position control.The data acquisition and reconstruction system is based on ATCA technology and consists of one acquisition board with integrated FPGA capabilities and a dual-core Intel module running RTAI Linux.In this paper, the tomographic algorithm and some preliminary results of the real-time plasma position control are presented with a performance benchmarking against other available positioning diagnostics. The algorithm has shown to be accurate and the system has successfully controlled the plasma position during a plasma current reversal.     

Effect of edge turbulent transport on scrape-off layer width on HL-2A tokamak

Effect of edge turbulent transport on scrape-off layer(SOL) width has been investigated in Ohmically heated L-mode plasma under limiter configurations on HL-2 A tokamak. It has been found that SOL width is doubled when plasma current decreases about 20%. With larger plasma current, E?×?B shear is stronger and has greater suppression effect on edge turbulent transport.SOL width is larger when power of relative density ?uctuation level in the edge region is larger.It is concluded that edge turbulent transport plays a significant role on SOL width. These experimental findings may provide a better understanding and controlling of power exhaust for present and future fusion devices.     

托卡马克等离子体杂质密度的光谱法测量研究

本文介绍了基于托卡马克等离子体被动光谱诊断获得杂质密度的方法。通过被动光谱诊断测量获得杂质线辐射的空间多道弦积分强度分布,利用强度标定系数转换为绝对光亮度分布;通过测量弦与等离子体位形,将弦积分的强度分布反演变换为径向体发射率。根据线辐射强度激发截面求出对应电离态的离子密度,最后采用杂质输运程序模拟计算得出总密度分布。以东方超环(Experimental Advanced Superconducting Tokamak,EAST)托卡马克装置上软X射线-极紫外光谱(Soft X-ray and Extreme Ultraviolet Spectrometers,XEUV)诊断测量到的Mo XXIX-Mo XXXII为例,描叙了获得Mo杂质密度分布的过程,获得的总误差小于10%。     

Smaller & Sooner: Exploiting High Magnetic Fields from New Superconductors for a More Attractive Fusion Energy Development Path

The current fusion energy development path, based on large volume moderate magnetic B field devices is proving to be slow and expensive. A modest development effort in exploiting new superconductor magnet technology development, and accompanying plasma physics research at high-B, could open up a viable and attractive path for fusion energy development. This path would feature smaller volume, fusion capable devices that could be built more quickly than low-to-moderate field designs based on conventional superconductors. Fusion’s worldwide development could be accelerated by using several small, flexible devices rather than relying solely on a single, very large device. These would be used to obtain the acknowledged science and technology knowledge necessary for fusion energy beyond achievement of high gain. Such a scenario would also permit the testing of multiple confinement configurations while distributing technical and scientific risk among smaller devices. Higher field and small size also allows operation away from well-known operational limits for plasma pressure, density and current. The advantages of this path have been long recognized—earlier US plans for burning plasma experiments (compact ignition tokamak, burning plasma experiment, fusion ignition research experiment) featured compact high-field designs, but these were necessarily pulsed due to the use of copper coils. Underpinning this new approach is the recent industrial maturity of high-temperature, high-field superconductor tapes that would offer a truly “game changing” opportunity for magnetic fusion when developed into large-scale coils. The superconductor tape form and higher operating temperatures also open up the possibility of demountable superconducting magnets in a fusion system, providing a modularity that vastly improves simplicity in the construction, maintenance, and upgrade of the coils and the internal nuclear engineering components required for fusion’s development. Our conclusion is that while tradeoffs exist in design choices, for example coil, cost and stress limits versus size, the potential physics and technology advantages of high-field superconductors are attractive and they should be vigorously pursued for magnetic fusion’s development.     

Continuous tokamak operation with an internal transformer

A large improvement in efficiency of current drive in a tokamak can be obtained using neutral beam injection to drive the current in a plasma which has low density and high resistivity. The current established under such conditions acts as the primary of a transformer to drive current in an ignited high-density plasma. In the context of a model of plasma confinement and fusion reactor costs, it is shown that such transformer action has substantial advantages over strict steady-state current drive. It is also shown that cycling plasma density and fusion power is essential for effective operation of an internal transformer cycle. Fusion power loading must be periodically reduced for intervals whose duration is comparable to the maximum of the particle confinement and thermal inertia time scales for plasma fueling and heating. The design of neutron absorption blankets which can tolerate reduced power loading for such short intervals is identified as a critical problem in the design of fusion power reactors.     

掺杂石墨在高能激光束和电子束作用下的热冲击行为

石墨被广泛用于当今的托卡马克装置中 ,作为真空室第一壁和偏滤器靶板的保护材料 ,也是未来聚变堆的一种候选面对等离子体材料。其抗化学溅射性能和抗热冲击性能受到广泛关注。用高能激光束和电子束轰击实验材料 ,模拟聚变堆面对等离子体材料在等离子体破裂时的工作状态 ,考察了 4种掺杂石墨材料在热冲击下的热腐蚀规律。实验结果表明 ,石墨掺杂能有效降低材料的烧蚀率。当激光单脉冲能量密度为 491 5KJ m2 时 ,冲击频率 1 0Hz,持续辐照 3 0秒后 ,几种掺杂石墨的失重率不超过2 1 3 6mg cm2 ,表现出了比纯石墨更优良的抗热冲击性能。     

Rview of pellet fueling

The injection of frozen pellets composed of the isotopes of hydrogen has become the leading candidate for refueling fusion power reactors based on the tokamak concept. This lofty position has been reached partly as a result of efforts to find an attractive solution to the perplexing problem of depositing atoms of fuel deep within the magnetically confined, hot plasma, and because of some recent experimental successes. To some extent, the relative merits of this technique will depend upon the distance that the cryogenic pellet will penetrate such a plasma, and the early exploratory research has addressed this problem on both theoretical and experimental fronts. The conclusion from the theoretical effort is that a protective blanket consisting of hydrogenic gas or cold plasma will envelope the pellet and partially shield the surface from the intense plasma heat flux. The blanket prolongs pellet lifetime, but penetration to the plasma center might require pellet injection velocities in excess of 10 km/s. The need for central penetration has not yet been established either theoretically or experimentally. The experiments performed to date have verified the existence of a shielding mechanism in general, and pellet ablation models that incorporate neutral gas shielding in particular are in adequate agreement with the experiments. Magnetic shielding effects are expected to contribute to, but not dominate, self-shielding in the higher plasma temperature regimes of the future. The tokamak plasma has demonstrated a surprising resilience even to massive density perturbations caused by the large refueling pellets used in present experiments. The characteristic discharge behavior is qualitatively not unlike that observed with gas puffing; but, for the first time, central plasma fueling has been studied, and this does not appear to be superior to refueling by partial pellet penetration. If relatively large pellets containing a significant fraction of the total plasma charge are acceptable in the present resistive plasma regimes, then it can be argued that they should have little impact on the gross stability of a hot thermonuclear tokamak plasma. Large pellets are preferable from the standpoint of attaining deep penetration, and this has important implications for the technology of pellet injection. The interesting velocity regime of 1 km/s has already been achieved with simple gun-type devices and this should be adequate for near-term tokamak experiments. Further improvements are anticipated, but the 10 km/s and above regime is uncertain; and, if current theory and experiments extrapolate to the future, such velocities might be desirable but unnecessary.     

Ignition Approach by Neutral Beam Injection Heating in Impurity Contaminated Tokamak Reactors

Available heating power by neutral beam injection in a tokamak reactor is evaluated semi-empirically. Using this estimated value, device and plasma parameters to ignite the plasma in impurity contaminated tokamak reactors are investigated. By lowering the plasma density and concurrently by enlarging the plasma minor radius or aspect ratio, the difficulty of NBI heating can be avoided, and the ignition is almost always possible both for trapped ion mode and Alcator scaling laws.     

Establishment and Assessment of Plasma Disruption and Warning Databases from EAST

Disruption database and disruption warning database of the EAST tokamak had been established by a disruption research group. The disruption database, based on Structured Query Language(SQL), comprises 41 disruption parameters, which include current quench characteristics, EFIT equilibrium characteristics, kinetic parameters, halo currents,and vertical motion. Presently most disruption databases are based on plasma experiments of non-superconducting tokamak devices. The purposes of the EAST database are to find disruption characteristics and disruption statistics to the fully superconducting tokamak EAST,to elucidate the physics underlying tokamak disruptions, to explore the influence of disruption on superconducting magnets and to extrapolate toward future burning plasma devices. In order to quantitatively assess the usefulness of various plasma parameters for predicting disruptions,a similar SQL database to Alcator C-Mod for EAST has been created by compiling values for a number of proposed disruption-relevant parameters sampled from all plasma discharges in the2015 campaign. The detailed statistic results and analysis of two databases on the EAST tokamak are presented.     

A Consideration on Increasing Current Density in Normal Conducting Toroidal Field Coil for Spherical Tokamak Power Plant

The center post is the most critical component as an inboard part of the toroidal field coil for the low aspect ratio tokamak. During the discharge it endures not only a tremendous ohmic heating owing to its carrying a rather high current but also a large nuclear heating and irradiation owing to the plasma operation. All the severe operating conditions, including the structure stress intensity and the stability of the structure, largely limit the maximum allowable current density. But in order to contain a very high dense plasma, it is hoped that the fusion power plant system can operate with a much high maximum magnetic field BT ≥12 T-15 T in the center post. A new method is presented in this paper to improve the maximum magnetic field up to 17 T and to investigate the possibility of the normal conducting center post to be used in the future fusion tokamak power plant.     

偏滤器物理设计与实验研究

应用B2-code模拟了偏滤器等离子体行为,优化了HL-2A装置偏滤器位形。研究了偏滤器刮削层中等离子体与器壁间过渡鞘层的离子碰撞效应,模拟研究了利用LHCD和NBI控制等离子体剖面分布在HL-2A中建立准稳态的反磁剪切位形。HL-2A装置首次实现了下单零点的偏滤器位形运行,完成了偏滤器初步物理实验,截至2004年底,获得等离子体电流320 kA,等离子体存在时间1 580 ms,环向磁场2.2 T。开展了高功率密度聚变堆偏滤器靶板的设计研究,特别是流动液态锂偏滤器靶板表面的物理过程的研究。探索性研究了用RF有质动力势改善偏滤器排灰效率和减少氚投料量。对FEB- E聚变堆偏滤器进行了优化设计。用电子束模拟对碳基材料及钨进行了高热负荷冲击实验,完成了钨/铜合金的热等静压焊接及热疲劳试验研究。研究了氦在钨中的滞留与热解吸行为。     

Application of Numerical Simulation in the Design of HECLIC Blanket for CFETR

China Fusion Engineering Test Reactor is a new tokamak device which is proposed by China National Integration Design Group. The fusion power is 50–200 MW and its plasma major radius and plasma minor radius are 5.7 and 1.6 m. The helium cooled lithium ceramic (HECLIC) blanket, as a key component of the tokamak, has the basic function to provide tritium breeding and plasma limiter. The blanket also provides main thermal and nuclear shielding of the vacuum vessel and ex-vessel components such as magnetic coils during plasma operations. With the development of the numerical simulation technology, more and more design parameters can be obtained by this method. Numerical simulation has been used for design and optimization, because some parameters are very hard to obtain though theoretical calculation. In this study, the simulation methods are investigated for HECLIC blanket design. Besides, design flow of the blanket is discussed and related analysis is also introduced to improve the design.     

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.     

Modeling of divertor power footprint widths on EAST by SOLPS5.0/B2.5-Eirene

The edge plasma code package SOLPS5.0 is employed to simulate the divertor power footprint widths of the experimental advanced superconducting tokamak(EAST)L-mode and ELM-free H-mode plasmas.The divertor power footprint widths,which consist of the scrape-off layer(SOL)widthλ_q and heat spreading 5,are important physical parameters for edge plasmas.In this work,a plasma current scan is implemented in the simulation to obtain the dependence of the divertor power footprint width on the plasma current I_p.Strong inverse scaling of the SOL width with I_p has been achieved for both L-mode and H-mode plasmas in the forms ofλ_(q,L-mode)=4.98×I_p~(-0.68)andλ_(q,H-mode)=1.86×I_p~(-1.08).Similar trends have also been demonstrated in the study of heat spreading with S_(L-mode)=1.95×I_p~(-0.542)and S_(H-mode)=0.756×I_p~(-0.872).In addition,studies on divertor peak heat load and the magnetic flux expansion factor show that both of them are proportional to plasma current.The simulation work here can act as a way to explore the power footprint widths of future tokamak fusion devices such as ITER and the China Fusion Engineering Test Reactor(CFETR).     

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.