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。     

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.     

可接受的聚变堆偏滤器方案第二部分：供选偏滤器的技术现状

聚变反应堆的安全运行要求排出５－３０ＭＷ．ｍ＾－２的热能量，所提出的不同的技术解决方案取决于不同的水冷铜散热器的设计，保护层可以用钨，也可以用铍或碳，这决定于与等离子体的相互作用。通过着重于与工业缺陷的配合进行的全面比较，实现了最优化设计。已提出了可靠的设计，并对有碳质护瓦的大部分在直至１０ＭＷ．Ｍ＾－１的功率水平下进行了娄千次循环的试验。     

偏滤器材料溅射过程的数值研究(英文)

为了估计一些偏滤器材料的腐蚀速率,应用一种描述溅射物理过程的新方法,基于离子输运的双群模型计算了一些偏滤器材料:钼、钨和锂在聚变带电粒子H+,D+,T+和He++的轰击下的溅射产额,并与 Monte-Carlo 方法计算得到的结果作了比较。这些结果对估计面对等离子体的部件寿命和分析聚变堆堆芯等离子体中的杂质水平估计是很有用的。     

HL-2M偏滤器热氮气烘烤仿真及优化

为提供高质量的等离子体真空运行环境,需对偏滤器进行高温烘烤。根据热传导与对流换热方程对偏滤器的烘烤过程进行了数值模拟及优化。结果表明:当热氮气等质量流量控制时,偏滤器回路压力损失逐渐增大,各部件烘烤温度爬升速率呈线性增加;当热氮气等体积流量控制时,偏滤器回路压力损失逐渐降低,各部件烘烤温度爬升呈线性增加。当初始条件近似相等(等质量流量为3×10~(-3)kg/s和等体积流量为4.8×10~(-4)m~3/s)时,前者的部件温升速率略低于后者,但各部件烘烤过程中最大温差均未超过90℃。     

聚变堆偏滤器冷却通道结构优化分析

偏滤器作为直接面向等离子体的内部部件之一,其表面承受的热流密度很高。为提高其冷却通道的冷却能力,降低此部件面向等离子体面边缘的温度,从改变冷却通道截面形状的角度提出了不同的改进方案,并采用理论计算与有限元仿真对原始设计和改进方案进行了流体、热和结构分析。结果表明:在冷却通道的横截面积不变的情况下,随着湿周周长的增加,冷却能力有所提高,钨边缘的局部温度过高得到改善;但冷却通道形状的变化出现应力集中现象,通过提高长宽比可适当提高其在材料应力限值下所能承受的稳态运行的热流密度。这些优化分析结果可为聚变堆偏滤器冷却结构的设计提供理论参考。     

用于保护托卡马克第一壁和偏滤器的厚碳化硼涂层

本文评述了用于保护受到高能热通量的托卡马克内表面的候选材料的各种碳化硼涂层的特性。这样的涂层可用各种方法生成：借助于氯化物和氟化物技术的化学气相沉积、气体转换、等离子体喷涂及反应－烧结。与纯碳材料相反，B4C具有低得多的化学和高温溅射，能够吸附氧和降低氢再循环。与薄的硼化膜相比，厚涂层能经受住象托卡马克偏滤器中那样的高热通量。比较分析表明，由扩散法（比如氟化物CVD和气体转换）形成的涂层更加能耐受热负载，而且最有前途的候选涂层之一是氟化物CVD涂层。     

Numerical Study of High Heat Flux Performances of Flat-Tile Divertor Mock-ups with Hypervapotron Cooling Concept

The hypervapotron (HV), as an enhanced heat transfer technique, will be used for ITER divertor components in the dome region as well as the enhanced heat flux first wall panels. W-Cu brazing technology has been developed at SWIP (Southwestern Institute of Physics), and one W/CuCrZr/316LN component of 450 mm×52 mm×166 mm with HV cooling channels will be fabricated for high heat flux (HHF) tests. Before that a relevant analysis was carried out to optimize the structure of divertor component elements. ANSYS-CFX was used in CFD analysis and ABAQUS was adopted for thermal–mechanical calculations. Commercial code FE-SAFE was adopted to compute the fatigue life of the component. The tile size, thickness of tungsten tiles and the slit width among tungsten tiles were optimized and its HHF performances under International Thermonuclear Experimental Reactor (ITER) loading conditions were simulated. One brand new tokamak HL-2M with advanced divertor con?guration is under construction in SWIP, where ITER-like ?at-tile divertor components are adopted. This optimized design is expected to supply valuable data for HL-2M tokamak.     

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.     

Study of Striated Heat Flux on EAST Divertor Plates Induced by LHW Using Infrared Camera

An upgraded infrared （IR） imaging system which provides a wide field of view （FOV） has been installed on the Experimental Advanced Superconducting Tokamak （EAST） to monitor the surface temperatures on plasma facing components. Modified magnetic topology induced by lower hybrid wave （LHW） can lead to the formation of striated heat flux （SHF} on divertor plates which can be clearly observed by IR camera. In this paper, LHW power modulation is applied to analyze the appearance of SHF. It is also demonstrated that deuterium （D） pellet injection and supersonic molecular beam injection （SMBI） can to some extent reduce the heat flux on the outer strike point （OSP）, but enhance the SHF on lower outer plates （LOP） of divertor. This may provide an optional approach to actively control the distribution of heat flux on diveror plates, which can protect materials from long duration high-heat flux.     

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.     

The Corrected Simulation Method of Critical Heat Flux Prediction for Water-Cooled Divertor Based on Euler Homogeneous Model

An accurate critical heat flux(CHF) prediction method is the key factor for realizing the steady-state operation of a water-cooled divertor that works under one-sided high heating flux conditions.An improved CHF prediction method based on Euler's homogeneous model for flow boiling combined with realizable k-ε model for single-phase flow is adopted in this paper in which time relaxation coefficients are corrected by the Hertz-Knudsen formula in order to improve the calculation accuracy of vapor-liquid conversion efficiency under high heating flux conditions.Moreover,local large differences of liquid physical properties due to the extreme nonuniform heating flux on cooling wall along the circumference direction are revised by formula IAPWSIF97.Therefore,this method can improve the calculation accuracy of heat and mass transfer between liquid phase and vapor phase in a CHF prediction simulation of water-cooled divertors under the one-sided high heating condition.An experimental example is simulated based on the improved and the uncorrected methods.The simulation results,such as temperature,void fraction and heat transfer coefficient,are analyzed to achieve the CHF prediction.The results show that the maximum error of CHF based on the improved method is 23.7%,while that of CHF based on uncorrected method is up to 188%,as compared with the experiment results of Ref.[12].Finally,this method is verified by comparison with the experimental data obtained by International Thermonuclear Experimental Reactor(ITER),with a maximum error of 6% only.This method provides an efficient tool for the CHF prediction of water-cooled divertors.     

Thermal Response of Bonded CFC/OFHC Divertor Mock-Ups for Fusion Experimental Reactors under Large Numbers of Cyclic High Heat Loads

Plasma facing components for fusion experimental reactors such as the ITER/FER will be exposed to severe heat loads under high heat flux and large number of thermal cycles. From the engineering point of view, divertor mock-ups with different armor tile materials have been prepared in order to investigate their overall performances, in particular an adhesive property between the armor tile and the heat sink metal. Thermal cycling tests of the divertor mock-ups have been carried out under ITER/FER relevant heat flux conditions in a particle beam engineering facility at JAERI. As results of the tests, it has been confirmed that boned carbon-fiber-composite/copper (CFC/OFHC) divertor mock-ups have resisted to 10.0 MW/m² for 1,000 cycles without cracks. At this experimental condition, the integrity and the durability of the bonds have also been confirmed. Furthermore, some bonded CFC/OFHC samples have resisted to 12.5 MW/m² for 1,000 cycles without increase of the surface temperature, although a small crack was observed at a corner of the bonded layer. Residual stress from brazing has also been analyzed for three-dimensional models. The analytical results were not different in the results of manufactured test samples that no cracks or detachments in many samples were observed.     

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.     

Design Comparison of Coated Structures for High Heat Flux Components in Fusion Reactor

High heat flux components in fusion reactors are examined from a viewpoint of structural design. Maximum admissible steady heat flux, which can be absorbed and removed by coated structures, was determined by one-dimensional numerical analysis basing on ASME code Sec. HI. Comparison of the current candidate materials is made in order to make heat flux as high as possible. The following conclusions were obtained. (1) Be-Cu structure can be used in order to remove high heat flux beyond 10 MW/m², however, the strong chemical activity of Be gives rise to a problem. (2) SiC-HT-9 and SiC-V alloy structures are promising in case of high heat flux less than 2 MW/m². (3) From thermal stress analysis tungsten and graphite are excellent coating materials and are available for many structural materials. The study of thermal shock resistance and thermal fatigue of these materials is now a problem to be solved.     

Design and Related Experiments of JT-60 Divertor Coils

The JT-60 divertor coils produce a separatrix configuration in divertor operations of JT-60. A suitable separatrix configuration was obtained for a plasma current of 2.1 MA with coil ampere turns of ± 0.755 MAT. A high primary membrane stress of 52 MPa was permissible at the welded joints of the copper conductor made on the site. The mechanical strength of the joints welded in a factory was also improved by means of a press treatment. Electric insulation materials were selected considering a degradation of with stand voltage characteristics due to high cyclic mechanical strain. Vacuum-tight coil cases were composed of rigid rings and U-shaped bellows made of Inconel-625 alloy, and designed to withstand plasma disruption with a current decay time constant of 3 ms. The maximum temperature of the conductor in the periodic operation of divertor discharges was below 155°C which was the allowable temperature of the coil insulation. Molybdenum armor plates coated with titanium carbide and Inconel-625 bellows cover plates were attached against high heat flux from plasma. Thermal and mechanical load tests were carried out using component models to evaluate their performance in advance of the final fabrication of the actual coils. The satisfactory performance of the divertor coils were demonstrated in the pre-operational power test.