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.     

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.     

Predictions for EAST Divertor Performance

A detailed study of the divertor performance in EAST has been performed for both its double null （DN） and single null （SN） configurations. The results of application of the SOLPS （B2-Eirene） code package to the analysis of the EAST divertor are summarized. In this work, we concentrate on the effects of increased geometrical closure and of magnetic topology variation on the scrape-off layer （SOL） and divertor plasma behavior. The results of numerical predictions for the EAST divertor operational window are also described in this paper. A simple Core-SOL- Divertor （C-S-D） model was applied to investigate the possibility of extending plasma operational space of low hybrid current drive （LHCD） experiments for EAST.     

Predictive Modeling of EAST Divertor Target Power Loading with Varying Chemical Sputtering Conditions

The Experimental Advanced Superconducting Tokamak （EAST） is being built in China to achieve high power and long pulse operation for studies of reactor-relevant issues under steady-state conditions. A major concern for EAST is the power handling capability of the divertor target plates, which is a critical issue for future high-powered steady-state tokamaks, such as ITER. Detailed modeling using B2/EIRENE code package and the most recent chemical sputtering data shows that the presence of strong chemical sputtering at the main chamber wall leads to strong carbon radiation in the periphery of the confined plasma, significantly reducing the heat fluxes to the target plates and facilitating plasma detachment at a lower density desired for lower hybrid current drive in EAST, with only a slight increase in Zeff at the edge. The target heat load can be further reduced by operating with a double-null divertor configuration, which also leads to a significant reduction in the edge Zeff. However, the code predicts that the double-null operation would result in a strong divertor asymmetry in target power loading, favoring the outside targets.     

Simulation Analysis of Divertor Performance in EAST

A detailed study of the divertor performance in the EAST has been conducted for both its double null and single null configurations. The results of the application of the SOLPS （B2/Eirene） code package to the analysis of the EAST divertor are summarized. Here we concentrate on the effects of the increased geometrical closure and variation in the magnetic topology on the behavior of divertor plasmas. The results of numerical predictions for the EAST divertor＇s operational window are also described in this paper.     

Review of Divertor Studies in LHD

In the Large Helical Device （LHD）, two different divertor configurations, i.e. helical divertor （HD） and local island divertor （LID）, are utilized to control the edge plasma. The HD with two X-points is an intrinsic divertor for heliotron devices, accompanied with a relatively thick ergodic layer outside the confinement region. Edge and divertor plasma behavior from low density to high density regimes is presented, referring to the divertor detachment. The effect of the ergodic layer on the edge transport is also discussed. On the other hand, the LID is an advanced divertor concept which realizes a high pumping efficiency by the combination of an externally induced magnetic island and a closed pumping system. Experimental results to confirm the fundamental divertor performance of the LID are presented.     

Investigation of Divertor Detachment in EAST by ＇Two-Point＇ Model

Divertor plasma detachment offers one of the most promising operating modes for fusion devices because of low target power loading. In this article a ＇two-point＇ model is used to investigate the formation of detachment and explore the route to detachment in EAST, in order to find an ideal operation window. The simulation results show that impurity radiation and ionneutral friction are the main causes of divertor plasma detachment at the target plates. Raising the safety factor and reducing the upstream power density provide effective means to achieve the detachment due to the increased radiation power fraction. Puffing Ar and Ne impurities and raising the safety factor can bring the upstream high plasma temperature region （above 100 eV） and the low target plasma temperature region （below 10 eV） close to each other in terms of the separatrix density. But it is difficult to find a common operating region which satisfies both conditions. High recycling and detached regimes provides an ideal operation window because of the steady upstream condition and low target power load.     

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.     

Thermal Fatigue Study on the Divertor Plate Materials

Thermal fatigue property of the divertor plate is one of the key issues that governs the lifetime of the divertor plate.Taking tungsten as surface material,a small-mock-up divertor plate was made by hot isostatic press welding (HIP),A thermal cycling experiment for divertor mock-up was carried out in the vacuum,where a high-heat-flux electronic gun was used as the thermal source,A cyclic heat flux of 9MW/m^2 was loaded onto the mock-up,a heating duration of 20s was selcted,the cooling water flow rate was 80ml/s.After 1000 Cycles,the surface and the W/Cu joint of the mock-up did not show any damage,The SEM was used to analyze the microstructure of the welding joint,where no cracks were found also.     

Structural Modification and Thermal and Structural Analysis of the Divertor in HL-2A

The re-design of the adoptable structure and the cooling manner of the divertor in the HL-2A tokamak is based on the parameters confirmed by the optimum divertor configuration and the primary modification scheme. The characteristics of the new divertor system include the double shear joint design on the domes and the outer target plates as well as the poloidal flow with toroidal manifolds. The results of the thermal and structural analysis of the outer target plates show that the design of the poloidal flow with toroidal manifolds can improve the capability of the target plates to withstand the heat loads, and the double shear joint design is compatible with the stress intensity requirements by the electromagnetic loads due to halo currents.     

自由表面液态锂偏滤器靶板物理过程研究

本文建立了一种高温液态锂蒸发、锂蒸气云等离子体运动、它对入射等离子体粒子屏蔽和锂蒸气云等离子体内的光子辐射和输运的综合物理模型.导出了与温度相关的蒸发功率.研究了静态液态锂表面在10 MJ/m2,1 ms高脉冲表面热负荷作用下考虑蒸发和不考虑蒸发两种情况下靶板温升并作了比较.结果表明定常自由表面液态锂靶板也可以取出大量表面热负荷.最后计算了入射到偏滤器靶室中的高能α粒子和弱相对论性电子在锂蒸气云等离子体中的能量沉积.     

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.     

偏滤器高热负荷材料研究

用热等静压焊接的方法对两种性能差异很大的金属材料钨和铜进行了成功的焊接。用SEM对断口和焊接界面在各种不同条件下焊接的性能作了对比分析。给出了两种样品的断裂特征和焊接过程中元素的扩散特点。钨和铜合金的结合主要是物理结合 ,是在高温高压下 ,材料表面微观的凹凸不平而产生的犬齿交合结合在一起的 ,扩散结合只占很少的部分。偏滤器靶板的寿命主要决定于靶板材料的热疲劳性能。在真空室中用大功率电子束作为热源进行了热疲劳试验。电子束的功率密度选为 9MW/m2 ,循环周期为 4 0s,冷却水流量为 80mL/s。用直径为 0 .3mm的NiCr NiSi热电偶测量了下材料表面的温度分布。结果发现 ,在冷却充分的情况下 ,表面最高温度约 4 0 0℃ ,钨铜焊缝处的平均最高温度约1 5 0℃。经过 1 0 0 0次的循环加热后 ,没有发现靶板材料出现破坏现象。对试验条件下的材料表面温度分布进行了计算机模拟计算。计算结果和试验测得的结果是相吻合的 ,表明试验结果是真实可靠的     

The Potential for Advanced Limiter/Divertor Systems

The Advanced Limiter-divertor Plasrna-facing Systems (ALPS) program was initiated in FY 1998 in order to evaluate the potential for improved performance and lifetime for plasma-facing systems. The main goal of the program is; to demonstrate the advantages of advanced limiter/divertor systems over conventional systems in terms of power density capability, component lifetime, and power conversion efficiency, while providing for safe operation and minimizing impurity concerns for the plasma.     

Impurity Measurement and Study on HL-2A Divertor Tokamak

HL-2A tokamak with two close divertors has been operated since 2003. In the experimental campaign of 2004 the divertor configuration has been successfully formed and the sillconization as a wall conditioning has been firstly done in this device. The divertor configuration can be reconstructed by the CFc code. Impurity behavior has been investigated during the experiment with divertor configuration and wall conditioning. The reduction of impurity is clear under both conditions of divertor configuration and siliconization.     

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.     

Experimental Study of the Atomic and Molecular Processes Related to Plasma Detachment in Steady-State Divertor Simulator MAP-II

Atomic and molecular processes relevant to the volumetric recombination phenomena were investigated in a linear divertor plasma simulator MAP-II. Volumetric recombination is induced in He plasma by puffing of He or H2. In the He puffing case, the reduction of the ion flux is dominated by the electron-ion recombination. In the H2 puffing case, however, it is dominated by the molecule-assisted recombination (MAR), which is characterized by the disappearance of the Helium Rydberg spectra and by the existence of the hydrogen negative ions. Current achievement and the future prospect are described.     

Perpendicular Conductivity and Self-Consistent Electric Fields in Small Size Divertor Tokamak Plasma Edge

The radial electric field of small size divertor tokamak in the vicinity of separtrix is simulated by using B2-SOLPS5.0 2D code, in which the most complete system of transport equations (Rozhansky et al., Nucl Fusion, 41:4, 2001) is solved including all the important perpendicular current and E × B drifts. Simulations demonstrated the following results: (a) It is shown that in the vicinity of the separatrix, the radial potential profile is determined by perpendicular currents (b) since, due to the pressure asymmetry, radial diamagnetic current integrated over the closed flux surface is not automatically zero, additional radial currents balance the diamagnetic current and make the average net current zero. (c) On the closed flux surfaces far from separetrix, where the pressure is almost constant, the calculated parallel currents (toroidal current) agree with Pfirsch–Schlueter currents.     

Simulation of Small Size Divertor Tokamak Plasma Edge Including Self-Consistent Electric Fields

The B2.SOLPES.0.5.2D code (Braams, Contrib Plasma Phys 36:276, 1996; Rozhansky and Tendler, Rev Plasma Phys 19:147, 1996) is applied for modeling SOL (Scrape off Layer) plasma in the small size divertor tokamak. Detailed distributions of the plasma heat flux and other plasma parameters in SOL, especially at the target plate of the divertor are found by modeling. The modeling results show that most of the electron heat flux and small part of ion heat flux arrive at target plate of the divertor, while, a large part of the ion heat flux and part of electron heat flux arrive at the outer wall. Also analysis of the role of poloidal E × B drifts in the redistribution of edge plasma is fulfilled.