Radiative divertor behavior and physics in Ar seeded plasma on EAST

To investigate the radiative divertor behavior and physics for the scenario of impurity seeded plasma in ITER, the radiative divertor experiments with argon(Ar) seeding under ITER-like tungsten divertor condition were carried out during recent EAST campaigns. The experimental results reveal the high efficiency of reducing heat load and particle flux onto the divertor targets owing to increased radiation by Ar seeding. We achieve detached plasmas in these experiments. The inner–outer divertor asymmetry reduces after Ar seeding. Impurities, such as Ar, C, Li, and W, exist in the entire space of the vacuum chamber during EAST operations, and play important roles in power exhausting and accelerating the plasma detachment process. It is remarkable that the contamination of the core plasma is observed using Ar seeding owing to the sputtering of plasma facing components(PFCs), particularly when Ar impurity is injected from the upper tungsten divertor.     

Study of multiple impurity seeding effect using SONIC integrated divertor code for JT-60SA plasma prediction

In order to study the potential impurity seeding operation regime of the future fusion devices, the first application of the integrated divertor code SONIC to the Ar + Ne mixed-impurity seeding operation of JT-60SA steady-state high-β plasma has been carried out. In the case, Ne is added to Ar-only seeding, the separatrix electron density has fell into the desired low separatrix electron density of the scenario. This is mainly because the D⁺ flow velocity towards the inner divertor has been increased by the Ne seeding. The resultant friction force transports Ar impurities towards the inner divertor region, while impurities are stagnated in the top of scrape-off layer (SOL) in the Ar-only seeding case. The higher impurity radiation power in the divertor regions and lower one in the SOL region above the X point have been obtained in mixed-impurity seeding cases, which show similar tendency as the Ar + Ne mixed-impurity seeding experiment in JT-60 U. At the core edge, Z_eff has been slightly increased and the radiation power has been decreased as the Ne seeding rate increases. The core plasma/impurity transport has been also evaluated by the TOPICS code using the impurity density at the core edge computed by the SONIC as a boundary parameter. The results show lower Z_eff and radiation power, and higher electron temperature in the core in the mixed-impurity seeding cases. Above possible contributors to the better energy confinement indicate that the mixed-impurity seeding operation might be more effective than Ar-only seeding operation.     

HL-2M装置常规偏滤器与雪花减偏滤器热工水力分析对比

采用SOLPS程序模拟预测HL-2M装置常规和雪花减偏滤器靶板上的热通量。当流入边缘等离子体区域的热功率约为10MW时,利用CFX/ANSYS软件分析这两类偏滤器各结构、冷却水温度分布及形变和热应力分布情况。结果表明:等离子体总功率相同,雪花减偏滤器靶板上的最高温度比常规偏滤器低169℃;雪花减偏滤器结构所承受的最大热应力和形变比常规偏滤器低约3/7。不改变热负载剖面分布,按一定比例提升热流密度或延长放电时间,雪花减偏滤器体现出比常规偏滤器靶板温升低、冷却水温均衡等优点。因此,雪花减偏滤器能处理更多流进偏滤器区域的热能,有效地降低偏滤器工程设计要求。     

Analyses of edge plasma characteristics in HL-2A

The edge plasma characteristics are studied by both a movable array of Mach/Reynolds stress/Langmuir 10-probes in the boundary region and the fixed flush probe arrays on the 4 divertor neutralization plates at the same toroidal cross-section in the HL-2A tokamak. The dependence of the Reynolds stress on poloidal flow in the edge plasma is analysed. The result indicates that the sheared poloidal flow in tokamak plasma can be induced by the radial gradient of Reynolds stress. In the divertor experiments of HL-2A, the profiles of the electron temperature, density and floating potential on divertor plates are measured by the flush probe arrays. The edge electron temperature in divertor configuration is higher than that in limiter configuration. The temperature asymmetry between outer and inner target plates is observed. The result of magnetic surface reconstructed from 18 Mirnov coils signals is presented. Both the particle recycling and the impurity flux in the bulk plasma during divertor discharges are discussed. Neutral gas pressure in divertor chamber, measured by fast ionization gauge during divertor discharge, is given.     

偏滤器等离子体中杂质对钨壁材料的侵蚀模拟研究

偏滤器是托卡马克中与等离子体直接接触的部件,为了保证装置的寿命,需要尽可能地减小等离子体对偏滤器靶板的侵蚀.本文用粒子模拟的方法研究了不同等离子体温度情况下碳和铍两种杂质离子对钨偏滤器侵蚀速率的影响.模拟首先得到稳定的鞘层结构、入射到靶板的离子流和能流密度,并通过统计获得了入射离子的能量和角度分布,最终根据这些物理参量,采用经验公式计算出钨靶板的侵蚀速率.研究表明,在等离子体温度不太高的情况下,钨靶板的热侵蚀几乎不起作用,而由于杂质离子对钨的物理溅射阈值较低,并且会通过鞘层加速获得能量,因此其对钨壁材料的物理溅射是导致靶板侵蚀的主要原因,另外靶板材料的侵蚀速率随着等离子体温度升高以及杂质含量增大而急剧增大.     

腐蚀和沉积对偏滤器靶板寿命的影响

讨论了偏滤器铍靶在聚变堆工况下的腐蚀问题，被物理溅射和辐射增强升华等过程腐蚀了的铍原子经电离后，部分将沉积到铍靶上，分析计算的结果表明，铍靶材被腐蚀后的自沉积将大大降低铍靶的净腐蚀率，在此基础，研究了一定浓度的硼杂质的沉积对铍靶的腐蚀率的影响，结果发现，铍靶前很低浓度的硼杂质即可大大降低了铍靶的腐蚀率，并且铍靶的腐蚀率随硼杂质浓度的增加而减少，这对聚变堆的稳态运行是有利的。     

Study on divertor plasma behavior through sweeping strike point in new lower divertor on EAST

A series of L-mode discharges have been conducted in the new ‘corner slot’ divertor on the Experimental Advanced Superconducting Tokamak (EAST) to study the divertor plasma behavior through sweeping strike point. The plasma control system controls the strike point sweeping from the horizontal target to the vertical target through poloidal field coils, with keeping the main plasma stability. The surface temperature of the divertor target cools down as the strike point moves away, indicating that sweeping strike point mitigates the heat load. To avoid the negative effect of probe tip damage, a method based on sweeping strike point is used to get the normalized profile and study the decay length of particle and heat flux on the divertor target λ_js, λ_q. In the discharges with high radio-frequency (RF) heating power, electron temperature T_e is lower and λ_js is larger when the strike point locates on the horizontal target compared to the vertical target, probably due to the corner effect. In the Ohmic discharges, λ_js, λ_q are much larger compared to the discharges with high RF heating power, which may be attributed to lower edge T_e.     

EAST辐射偏滤器充气系统改造及初步实验结果

介绍了EAST偏滤器充气系统的改进措施,压电阀与冷屏一起安装在EAST颈管中,以缩短充气管道,将充气系统的延迟时间减少约200ms。冷屏起冷却阀门并屏蔽电磁场的作用。讨论了在准稳态条件下,延迟时间给偏滤器热流反馈控制带来的负面效应。在中性束加热条件下,通过在外靶板注入Ar/D2(1:4)混合气体研究了靶板离子饱和流和热流分布的变化及充气对主等离子体的影响,得到了靶板离子饱和流显著降低、热流减少的结果。     

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介绍了EAST偏滤器充气系统的改进措施,压电阀与冷屏一起安装在EAST颈管中,以缩短充气管道,将充气系统的延迟时间减少约200ms。冷屏起冷却阀门并屏蔽电磁场的作用。讨论了在准稳态条件下,延迟时间给偏滤器热流反馈控制带来的负面效应。在中性束加热条件下,通过在外靶板注入 Ar/D2(1：4)混合气体研究了靶板离子饱和流和热流分布的变化及充气对主等离子体的影响,得到了靶板离子饱和流显著降低、热流减少的结果。     

Numerical Modelling of Impurity Retention in the ITER Tokamak Scrape-off Layer Plasma

The hydrogen plasma and carbon ions flows generated at the target plates in the ITER tokamak scrape-off layer are numerically investigated. A 2D model of hydrogen plasma and impurity ions is presented. The model is based on the electron-ion Braginski fluid equations [1] for hydrogen plasma and rate equations for impurities. Arbitrary level of impurity ions concentration is assumed. Recycling of hydrogen, sputtering and self sputtering of carbon atoms at the target plates are taken into account. Equations of the model are solved in the slab geometry using 2D-multifluid numerical code EPIT [2]. Problems of impurity ions retention and radiation in the divertor volume are analyzed. Results of calculations for ITER tokamak boundary plasma are presented, showing that poor retention is likely at high impurity concentration in the divertor volume. The radiation power can be a significant part of ingoing energy.     

FEB偏滤器等离子体温度及靶板运行温度的优化

本文以石墨偏滤器靶板材料为例，用喷气或杂质注入对偏滤器靶板前等离子体温度进行控制，要刻薄虎器等离子体的各种变标关系及石墨靶板被腐蚀的程度、优化了ＦＥＢ偏滤器靶板前等离子体温度和靶板运行温度。     

FEB-E动态气靶偏滤器的研究

在ＦＥＢ－Ｅ设计阶段,偏波器从开放式固定板靶优化为封闭式气体靶,以改善偏滤器的杂质控制和增加离子与气体的相互作用,通过喷气和注入杂质获得的部分脱靶等离子体形成了动态气体靶,喷气能降低删削层（ＳＯＬ）处等离子体温度,沪入的杂质增加了ＳＯＬ处的辐射功率,使靶板的负载降低,用ＮＥＷＴ１Ｄ编码模拟了ＳＯＬ处等离子体和杂质（硼杂质）的输运,得到了杂质、等离本温度和等离了体密度分布。着眼于杂质的滞留物辐射,优     

ITER偏滤器钨/铜穿管模块的高热负荷测试及事后分析

对ITER偏滤器钨/铜穿管模块进行了高热负荷测试。测试结果表明,所有的模块都成功承受住了热负荷测试并满足了IO的要求。在高热负荷测试后,对模块进行了事后分析。发现了钨的纵向开裂和Cu/CuCrZr界面的开焊现象。断口的SEM图像表明,钨的开裂方式为脆性的沿晶开裂。通过金相学分析,发现了钨的再结晶现象和钨表面的再结晶深度。还通过有限元分析软件模拟得到了模块的温度分布,并且与实验观察的再结晶现象能够很好的对应。     

感应耦合等离子体增强射频磁控溅射沉积ZrN薄膜及其性能研究

利用感应耦合等离子体(ICP)增强射频磁控溅射技术在Si(111)片和M2钢表面制备了ZrN薄膜,研究了基片的温度和ICP功率对ZrN薄膜的结构以及性能影响.研究发现：在基片温度≤300℃沉积的ZrN薄膜择优取向为(111);基片温度达到450℃时薄膜出现ZrN(200)衍射峰,ZrN(111)晶面的织构系数明显降低.传统磁控溅射沉积薄膜为柱状结构,当ICP为200 W,基片温度为300℃时沉积薄膜中柱状晶体消失;随着基片温度的升高,N/Zr元素比例降低,并且薄膜的电阻率下降;相对于传统溅射,ICP增强射 关键词： 感应耦合等离子体 磁控溅射 ZrN 微结构     

东方超环托卡马克高约束模式边界等离子体输运数值模拟研究

利用SOLPS5.0模拟研究东方超环托卡马克（EAST）高约束模式时的刮削层等离子体. 在高约束模式放电实验参数（第36291炮）的限制下,通过调整上游区径向反常输运系数来实现高约束模式模拟,在上游电子密度和温度与实验符合的条件下能够很好地进行下游区模拟. 在实现高约束模拟的基础上又分别研究了漂移项对偏滤器靶板能流不对称性的影响和上游能流衰减宽度对靶板能流密度峰值的影响. 通过模拟发现,漂移是导致EAST放电内外靶板不对称性的主要原因,增大上游能流衰减宽度可以明显降低入射到靶板的峰值热流,并且偏滤器区辐射以及与中性粒子的相互作用减小了能流的衰减宽度对达到靶板能流的影响. 关键词： 托卡马克 高约束模式 SOLPS5.0 漂移     

Monte Carlo Simulation of Helium Transport in the Vicinity of Divertor Plate

A Monte Carlo code “1DHET” has been utilized for simulating the transport of neutral helium near divertor flat pumping target plate. The dominant atomic processes of electron impact ionization and elastic scattering by plasma ions are included. The thermal and streaming motion of the ions along the magnetic field is taken into consideration. Calculations show that the probability of neutral helium turning back to the target plate will increase by 45% if the divertor edge plasma density is smoothly raised up to 4.7 times higher at the divertor plate than that at 7cm apart inward as compared with the ordinary outward decreasing density profile while keeping temperature profile fixed. The temperature decreases smoothly from 40 eV at z = 7 cm to 3 eV at the plate. The effects of angle between the magnetic field and the target plate, the temperature of plate, edge plasma sheath potential and different target materials on the probability of helium particles scattered back to the plate are studied.     

聚变堆熔盐冷却偏滤器靶板结构设计与分析

为提高偏滤器的抗中子辐照能力,兼顾高热承载能力和聚变堆经济性的需要,提出了基于熔盐冷却(MSC)的偏滤器靶板结构设计。它采用FLiNaK作为冷却剂,钨镧合金为热沉材料,钨为第一壁材料。通过数值计算评估了靶板的热负荷承载能力,并完成了偏滤器冷却剂回路设计,优化了偏滤器各模块之间的流量分配。此MSC偏滤器靶板设计可以有效去除10~15MW•m^-2热负荷,为适应未来聚变堆偏滤器靶板发展的需要提供了一种设计解决方案。     

Heat Flux on EAST Divertor Plate in H-mode with LHCD/LHCD+NBI

Based on the surface temperature measured by the infrared camera on the experimental advanced superconducting tokamak(EAST), the heat fluxes on the lower outer divertor target plate during H-mode with the lower-hybrid wave current drive(LHCD) only and with the LHCD combined with the neutral beam injection(NBI) are calculated by the DFLUX code and compared. The analyzed discharges are lower single null divertor configuration discharges. In the case with the LHCD only(I_p~400 kA, P_(LHCD)~2 MW), ELM-free appears after L-H transition with the peak heat flux on the lower outer target plate less than 1 MW/m2. However, there is no ELMfree appearing after the L-H transition in the case with the LHCD+NBI(I_P~300 kA, P_(LHCD)+P_(NBI)~2 MW).The results show that the peak heat fluxes on the lower outer target plate in the LHCD+NBI H-mode cases are larger than those in the LHCD H-mode under the similar auxiliary heating power. This is because the heat flux profiles of the lower outer target plate as a function of plate location in ELMing with the LHCD+NBI are narrower than those with the LHCD only. The results are consistent with the results in terms of the scrape-off layer width observed in the EAST.     

Neutral transport code for rovibrational population calculation of molecular hydrogen in large helical device plasmas

An rovibrationally resolved collisional-radiative model of molecular hydrogen (K. Sawada and M. Goto, Atoms 4: 29, 2016) is included in the authors' neutral transport code for large helical device (LHD) plasmas. The spatial distributions of the electron temperature and density, including the divertor and divertor leg regions, are given to the code. The molecules released from the graphite divertor target are tracked. The initial rotational and vibrational states of the released molecules are provided by simulation using a molecular dynamics model. The rovibrational population produced in an LHD plasma is evaluated.     

Evolution of the high-field-side radiation belts during the neon seeding plasma discharge in EAST tokamak

Divertor detachment achieved by injecting impurities or increasing density is always accompanied with various local radiation phenomena in the boundary or core plasma. This paper presents the formation and evolution of the high-field-side (HFS) radiation belts during the neon seeding plasma discharge in upper single null configuration with two directions of toroidal magnetic field in EAST tokamak. The neon mixed with deuterium seeding can induce the divertor detachment with strong radiation belts in the HFS scrape-off layer (SOL) region. With the increase of the radiation power, the plasma discharge will transit from H-mode to L-mode, and meanwhile the radiation belts move away from the near X-point to HFS SOL. When the radiation power is high enough, the radiation belts begin to move further to the other X-point along the HFS SOL, and even cause plasma disruption. The results indicate that the behavior of the radiation belts is related to the radiation power, plasma confinement performance and state of divertor detachment, which is useful for developing better feedback control methods to achieve high-performance radiative divertor operation mode.