Impurity Measured by VUV Spectrometer and OMA on HL-2A Tokamak

Impurity is one of the key issues on a great impact to the quality of tokamak plasma.HL-2A is the first divertor tokamak in China. In this paper the experimental results are presented on impurity through the line emission measurement in the campaign in 2003 under the limiter and divertor configurations. The low-Z impurities such as carbon and oxygen are the most important components in the plasma, but their content are not so high to affect the discharge quality. The high-Z impurities such as copper and ferrum are not essential. The emission intensity of impurity is clearly decreased during the divertor configuration formed.     

The Spectroscopic Systems for the Study of Light Impurity Particle Transport in the HT-7 Tokamak

Three spectroscopic systems have been developed for the study of light impurity particle transport in the HT-7 tokamak. A visible multi-channel spectroscopic system （VIS） is used to obtain the brightness distribution of the line emission from ionized light impurities. The profile of Zeff（r） has been obtained from the visible multi-channel bremsstrahlung measurement （VB）. The system with a rotating hexahedral mirror for space-time resolved spectroscopy measurement from ultraviolet to visible （UV） can provide the brightness distribution of two different emission lines of the light impurities simultaneously. The emissivities by these multi-channel measurements can be obtained by Abel inversion. The measurement was performed in typical OH discharges in the HT-7 tokamak. The carbon particle transport was analyzed. The feasibility of these diagnostic systems for the impurity particle transport study is clearly demonstrated.     

Divertor Experiments with MBI and Strong Gas Puffing on HL-2A

In the HL-2A 2004 experiment campaign, pulsed molecular beam injection （MBI） and strong hydrogen gas puffing under the divertor configuration were used for gas fueling. The experimental results show that the MBI of hydrogen can reduce the heat flux to the divertor target plate. The electron temperature measured by the Langmuir probe array decreases significantly during the injection of the molecular beam whereas the electron density increases. This indicates that the plasma pressure near the target plates tends to be constant at a new equilibrium level. In the divertor plasmas with strong hydrogen gas puffing a high plasma density up to 4.4 × 10^19 m^-3 was achieved. In addition, a phenomenon similar to the partially detached divertor regime was observed, which is being studied in open divertor tokamaks such as DIII-D to reduce the peak heat flux on the target plates near the separatrix. After a strong gas puffing the electron temperature measured on the outer divertor target plate near the separatrix decreases till below 5 eV or even lower, but that of the farther outer divertor target plate does not change obviously; and the CIII and the Ha emissions at the plasma edge decrease as expected, but the Ha emission near the Xpoint increases. These results reflects some interesting characteristics, which needs to be studied by further modeling and experiments.     

Imaging System and Plasma Imaging on HL-2A Tokamak

As a new diagnostic means, plasma-imaging system has been developed on the HL-2A tokamak, with a basic understanding of plasma discharge scenario of the entire torus, checking the plasma position and the clearance between the plasma and the first wall during discharge. The plasma imaging system consists of (1) color video camera, (2) observation window and turn mirror, (3) viewing & collecting optics, (4) video cable, (5) Video capture card as well as PC. This paper mainly describes the experimental arrangement, plasma imaging system and detailed part in the system, along with the experimental results. Real-time monitoring of plasma discharge process, particularly distinguishing limitor and divertor configuration, the imaging system has become key diagnostic means and laid the foundation for further physical experiment on the HL-2A tokamak.     

Numerical Simulation and Experimental Identification of Divertor Configuration in the HL-2A Tokamak

Single- and double-null divertor configurations in HL-2A are simulated by SWEQU equilibrium code. Lower divertor discharges in the first physics campaign have been achieved by two kinds of power supply method of multipole-field coils. Single-null divertor configuration has been identified by visible photography, target probe arrays and the reconstructed magnetic surface. Magnetic separatrix and minor radius of plasma column are obtained by a reconstructed code of multiple current filaments using 18 Mirnov signals.     

Flow Field and Thermal Analysis of the Divertor Target Plate for HL-2A Tokamak

In the initial phase of the physics experiment, the double-null divertor plates used consist of graphite armor tiles, Mo-alloy intermediate layers and Cu-alloy coolant tubes. In the later operating phase, tungsten will be used as armor tiles. A multi-physical field numerical analysis method is used in this paper. Its analysis model reflects more realistically the real divertor structure than other models. Two-dimensional （2D） and three-dimensional （3D） fluid flow field, temperature distribution and thermal stress analyses of the divertor plates are carried out by the ANSYS code. During the physics experimental phase with a heat flux of 1 MW/m2, a coolant velocity of 5.48 m/s, and a thermal stress of 750 kg/cm2, the graphite armor tiles successfully meet the requirements of temperature, thermal stress and sputtering erosion. The tungsten armor will be considered as a second candidate. The result of simulation can be used for upgrading the design parameters of the HL-2A poloidal divertor.     

Improvement of Plasma Performance with Lithium Wall Conditioning in Aditya Tokamak

Lithiumization of the vacuum vessel wall of the Aditya tokamak using a lithium rod exposed to glow discharge cleaning plasma has been done to understand its effect on plasma performance. After the Li-coating, an increment of ～100 eV in plasma electron temperature has been observed in most of the discharges compared to discharges without Li coating, and the shot reproducibility is considerably improved. Detailed studies of impurity behaviour and hydrogen recycling are made in the Li coated discharges by observing spectral lines of hydrogen, carbon, and oxygen in the visible region using optical fiber, an interference filter, and PMT based systems. A large reduction in O I signal (up to ～ 40% to 50%) and a 20% to 30% decrease of H α signal indicate significant reduction of wall recycling. Furthermore, VUV emissions from O V and Fe XV monitored by a grazing incidence monochromator also show the reduction. Lower Fe XV emission indicates the declined impurity penetration to the core plasma in the Li coated discharges. Significant increase of the particle and energy confinement times and the reduction of Z eff of the plasma certainly indicate the improved plasma parameters in the Aditya tokamak after lithium wall conditioning.     

Sawtooth Activities During Impurity Injection by Laser Blow-off in HL-2A

In order to study impurity transport and their progression in the plasma, small amount of high-Z impurities were injected into HL-2A ohmic heated plasmas by laser blow-off and traced with good spatial temporal resolution by a soft X-ray camera. An analysis of the evolution of the impurities＇ radiation distribution in the plasma revealed that impurity transport in the central region was profoundly different from that outside of it. It was found that the sawteeth tended to be inverted on the central soft X-ray chord signals after the impurity injection and the soft X- ray profiles exhibited discontinuities during the movement of the impurities in the plasma central region. Detailed 2D visualization of the evolution of impurities＇ radiation using computer tomography has provided further understanding of the properties of impurity transport in the HL-2A plasma.     

Infrared Imaging Bolometer for the HL-2A Tokamak

An infrared imaging bolometer diagnostic has been upgraded recently to be adapted for the complications of the signal-to-noise ratio arising from the low level of plasma radiation and high reflectivity of low energy photon(6.2 eV).It utilizes a platinum foil,blackened on both sides with graphite spray,as the bolometer detector.The advantage of the blackened foil is the light absorption extending into the infrared.After a careful calibration of the foil,the incident power density distribution on the foil is determined by solving the heat diffusion equation with a numerical technique.The local plasma radiated power density is reconstructed with a minimum fisher information regularization method by assuming plasma emission toroidal symmetry.Comparisons of the results and the profiles measured by an ordinary bolometric detector demonstrate that this method is good enough to provide the plasma radiated power pattern.The typical plasma radiated power density distribution before and after high mode(H-mode) transition is firstly reconstructed with the infrared imaging bolometer.Moreover,during supersonic molecular beam injection(SMBI),an enhanced radiation region is observed at the edge of the plasma.     

The Principle of Plasma Boundary Identification on HL-2A Tokamak

Plasma boundary identification is a basic task for studies on equilibrium and confinement in a divertor tokamak. With the progress on the experiments after engineering experiments, the boundary identification becomes an important issue for HL-2A. In order to satisfy the requirements of preciseness, simplified measurements and quickness, the filament current method instead of solving the equilibrium equations is used to identify plasma boundary on HL-2A. The involved principle, mathematics and the progresses, which have been made with this method, are given.     

The Impurity Cleaning Experiments on HL-1M Tokamak

The glow discharge cleaning(GDC) is widely used for the fusion devices such as JT-60U, JET and TFTR at present. The HL-1M Tokamak has a single shell of the vacuum chamber liner made of 00Cr18Ni10Ti and graphite. 6％ of the interior wall is covered by the graphite. Two sets of ultra-high vacuum pumping machines are used on HL-1M Tokamak. There are various impurities in the vacuum chamber without cleaning. The metal impurities Cr, Ni comes from the vacuum chamber material. The first wall processing methods such as the siliconization     

Impurity Measurements in Plasma Boundary of JIPP T-II Tokamak/Stellarator

Impurity transport in the scrape-off layer of the JIPP T-H Tokamak/Stellarator hybrid torus machine in Institute of Plasma Physics, Nagoya University has been investigated by a probe measurement. Silicon and Cu probes were inserted into the boundary plasma during discharges. The surface of the probes was analyzed by several analytical techniques including AES, XPS, RBS and PIXE. Main metallic impurities deposited on the probes were Mo and Fe which originated from the limiters and the inner wall of the vacuum chamber. Depth concentration profile of deposited impurities was measured by AES along with sputter-etching of the surface with Ar ions. Deposited Fe impurity had the maximum concentration at the top surface of the probe, while the depth profile of Mo showed the maximum concentration at around 8 A from the top surface. Transportation energy of these impurities are discussed on the basis of the projected range data.     

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.     

Analysis of Divertor Heat Flux with Infrared Thermography During Gas Fuelling in the HL-2A Tokamak

In HL-2A tokamaks, the behavior of heat flux deposited on the divertor targets has been studied during deuterium gas fuelling. The heat flux is reduced significantly after supersonic molecular beam injection （SMBI） fuelling during Ohmic and electron cyclotron resonance heating （ECRH） divertor discharges. The SMBI fuelling causes an increase in the plasma density and this change results in the experienced change of the edge properties. Most of this reduction in divertor target heat flux occurs together with a high plasma radiation region located at near the X-point. The largest reduction in heat flux profiles is observed at the outboard divertor separatrix strike point, while the heat flux far from the strike point remains almost unchanged. In particular, with SMBI multi-pulses gas fuelling, a partially detached divertor regime is observed with a highly radiating region at the X-point. With the onset of the partially detached divertor regime, a sudden drop in both heat flux and power flow on the divertor target is observed. The reduction in power load on the divertor targets is roughly equal to the increase in plasma radiation loss.     

Measurement of Turbulence Propagation Velocity Using Doppler Reflectometer on HL-2A Tokamak

O-mode Doppler reflectometer has been successfully developed as an important diagnostic system on HL-2A. It can be used to measure the turbulence propagation in both plasma edge and confinement zone. The Doppler reflectometer system consists of two fixed frequency homodyne receivers： 15 GHz （corresponding to cutoff density of 0.3×10^19 m^-3） and 33 GHz （corresponding to cutoff density of 1.35× 10^19 m^-3）. The Doppler reflectometry principle and the experimental arrangements on HL-2A are presented. Meanwhile, the experimental Doppler reflectometric spectra under different discharge conditions, with and without ECRH, were obtained. Furthermore, the turbulence propagation velocity change and the profile were also observed in different discharge conditions.     

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。     

Two-Dimension Investigation of HL-2A Tokmak Divertor Plasma

Recently, attention has been drawn to the particle and energy transport of the divertor plasma and Scrape-off Layer plasma in TOKAMAK. The crucial tasks faced by the divertor can be divided into four main aspects:     

Study of Carbon Impurity Transport at SOL in EAST

Transport of carbon in the edge plasma of EAST with 啊heating power Pin of 8 MW is studied using DIVIMP code. The background plasmas for DIVIMP are taken from the results by using B2.5-Eirene code. For different plasma densities at the core-SOL interface and the different divertor operational regimes, namely low recycling, high recycling and detachment, the simulated results show that the impurity density in SOL is higher for the high recycling regime than that for the low recycling regime, while impurity density in SOL is lower for the detachment regime than that for both the low and high recycling regimes.     

Study of Plasma MHD Equilibrium in HL-2A Tokamak

EFIT（Equilibrium Fitting） code is successfully transplanted for HL-2A tokamak parameters. The evolution of the equilibrium configurations for shot 2898 is simulated. It is shown that the discharge starts with a limiter configuration and develops into a divertor configuration gradually after t = 200 ms. The latter lasts until the end of the discharge at t = 900 ms. The evolution of the major plasma shape parameters such as the boundary magnetic fluxes, the positions of the x-point and magnetic axis, and the minor radii obtained are calculated and compared with the experimental results. The agreement between the simulation and experiments are shown to be reasonable. The possibility for discharge quality improvement is discussed.     

Vacuum System for HL-2A Tokamak

The vacuum system for HL-2A was built in 2003. The test results indicated that this system is feasible. It consists of three main parts: a pumping system, a pumping divertor and a glow discharge cleaning (GDC) system. For the pumping system, there are three main functions: (1) evacuating the vacuum vessel thus to produce an ultra high vacuum, (2) removal of impurities released during baking and (3) pumping during GDC. The pumping divertor controls the particles at the plasma edge and the GDC system provides a clean wall conditioning. During the first campaign of physical trial experiment on HL-2A, the ultimate pressure reached 4.6 × 10-6 Pa, and the total leakage and outgassing rate in 12 hours was 1.8 × 10-5 Pa·m3/s, which is close to that of ASDEX.