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 Ha 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 Zeff of the plasma certainly indicate the improved plasma parameters in the Aditya tokamak after lithium wall conditioning.

Development of a helicon-wave excited plasma facility with high magnetic field for plasma-wall interactions studies

The high magnetic field helicon experiment system is a helicon wave plasma（HWP）source device in a high axial magnetic field（B0）developed for plasma–wall interactions studies for fusion reactors.This HWP was realized at low pressure（5×10^-3-10 Pa）and a RF（radio frequency,13.56 MHz）power（maximum power of 2 k W）using an internal right helical antenna（5 cm in diameter by 18 cm long）with a maximum B0of 6300 G.Ar HWP with electron density～10^18–10^20m^-3 and electron temperature～4–7 e V was produced at high B0 of 5100 G,with an RF power of 1500 W.Maximum Ar^＋ion flux of 7.8×10^23m^-2s^-1 with a bright blue core plasma was obtained at a high B0 of 2700 G and an RF power of 1500 W without bias.Plasma energy and mass spectrometer studies indicate that Ar^＋ ion-beams of 40.1 eV are formed,which are supersonic（～3.1cs）.The effect of Ar HWP discharge cleaning on the wall conditioning are investigated by using the mass spectrometry.And the consequent plasma parameters will result in favorable wall conditioning with a removal rate of 1.1×10^24N2/m^2 h.

Study of the tungsten sputtering source suppression by wall conditionings in the EAST tokamak

The steady fusion plasma operation is constrained by tungsten(W)material sputtering issue in the EAST tokamak.In this work,the suppression of W sputtering source has been studied by advanced wall conditionings.It is also concluded that the W sputtering yield becomes more with increasing carbon(C)content in the main deuterium(D)plasma.In EAST,the integrated use of discharge cleanings and lithium(Li)coating has positive effects on the suppression of W sputtering source.In the plasma recovery experiments,it is suggested that the W intensity is reduced by approximately 60%with the help of~35 h Ion Cyclotron Radio Frequency Discharge Cleaning(ICRF-DC)and~40 g Li coating after vacuum failure.The first wall covered by Li film could be relieved from the bombardment of energetic particles,and the impurity in the vessel would be removed through the particle induced desorption and isotope exchange during the discharge cleanings.In general,the sputtering yield of W would decrease from the source,on the bias of the improvement of wall condition and the mitigation of plasmawall interaction process.It lays important base of the achievement of high-parameter and longpulse plasma operation in EAST.The experiences also would be constructive for us to promote the understanding of relevant physics and basis towards the ITER-like condition.

Kinetic simulation of an electronegative plasma with a cut-off distribution and modified Bohm criterion

A kinetic simulation model has been employed to study the properties of an electronegative magnetized plasma sheath assuming the cut-off distribution of electrons and negative ions.The fundamental kinetic Bohm condition for the electronegative magnetized plasma has paramount importance for the formation of a stationary plasma sheath near the material wall.The presence of an oblique magnetic field in an electronegative plasma affects the ion distribution at the plasma injection side and at the wall.The temperature profile of negative charged particles has a non-uniform distribution,which determines the energy flow towards the wall.

Line identification of boron and nitrogen emissions in extreme-and vacuumultraviolet wavelength ranges in the impurity powder dropping experiments of the Large Helical Device and its application to spectroscopic diagnostics

An impurity powder dropper was installed in the 21 st campaign of the Large Helical Device experiment(Oct.2019–Feb.2020)under a collaboration between the National Institute for Fusion Science and the Princeton Plasma Physics Laboratory for the purposes of real-time wall conditioning and edge plasma control.In order to assess the effective injection of the impurity powders,spectroscopic diagnostics were applied to observe line emission from the injected impurity.Thus,extreme-ultraviolet(EUV)and vacuum-ultraviolet(VUV)emission spectra were analyzed to summarize observable impurity lines with B and BN powder injection.Emission lines released from B and N ions were identified in the EUV wavelength range of 5–300Ameasured using two grazing incidence flat-field EUV spectrometers and in the VUV wavelength range of 300–2400Ameasured using three normal incidence 20 cm VUV spectrometers.BI–BV and NIII–NVII emission lines were identified in the discharges with the B and BN powder injection,respectively.Useful B and N emission lines which have large intensities and are isolated from other lines were successfully identified as follows:BI(1825.89,1826.40)A(blended),BII 1362.46A,BIII(677.00,677.14,677.16)A(blended),BIV 60.31A,BV 48.59A,NIII(989.79,991.51,991.58)A(blended),NIV765.15A,NV(209.27,209.31)A(blended),NVI 1896.80A,and NVII 24.78A.Applications of the line identifications to the advanced spectroscopic diagnostics were demonstrated,such as the vertical profile measurements for the BV and NVII lines using a space-resolved EUV spectrometer and the ion temperature measurement for the BII line using a normal incidence 3 m VUV spectrometer.

Preliminary Investigation of Impurity and Radiation Behaviors in the Startup Phase of HL-2A Tokamak

As the startup phase of HL-2A operation and the first divertor tokamak device in China, it is important to identify the species of impurity, estimate the concentration of impurities and compare the different behaviors of radiation in limiter and divertor configurationt. In startup phase of HL-2A, glowing discharge cleaning （GDC） is used as a conventional wall conditioning technology and titanium gettering is used occasionally in closed divertor chamber during SN configuration discharges phase.

Preliminary results of in situ laser-induced breakdown spectroscopy for the first wall diagnostics on EAST

Post-mortem methods cannot fulfill the requirement of monitoring the lifetime of the plasma facing components （PFC） and measuring the tritium inventory for the safety evaluation. Laserinduced breakdown spectroscopy （LIBS） is proposed as a promising method for the in situ study of fuel retention and impurity deposition in a tokamak. In this study, an in situ LIBS system was successfully established on EAST to investigate fuel retention and impurity deposition on the first wall without the need of removal tiles between plasma discharges. Spectral lines of D, H and impurities （Mo, Li, Si ） in laser-induced plasma were observed and identified within the wavelength range of 500-700 nm. Qualitative measurements such as thickness of the deposition layers, element depth profile and fuel retention on the wall are obtained by means of in situ LIBS. The results demonstrated the potential applications of LIBS for in situ characterization of fuel retention and co-deposition on the first wall of EAST.

Impurity behaviours of HL-2A divertor Tokamak during the initial operation period

The divertor configuration was successfully formed and the siliconization as a wall conditioning was first achieved on HL-2A tokamak experimentally in 2004. The divertor configuration is reconstructed by the use of the CFC code. Impurity as an important issue is investigated in the experiments with divertor configuration and wall conditioning. Impurities dramatically decrease after both the divertor configuration is formed and silicon is coated on the surface of the vacuum vessel.

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.

面向EAST装置长脉冲高参数等离子体运行的真空系统性能提升

托卡马克核聚变装置真空系统为等离子体放电真空腔室提供良好的真空环境、洁净器壁、实时燃料注入及粒子排出等条件,直接影响高温等离子体的品质。真空系统作为EAST装置的重要组成部分,包括真空抽气系统、等离子体加料系统、壁处理系统以及测量与控制系统。针对EAST装置长脉冲高参数等离子放电的需求,对真空系统进行了一系列的升级改造,包括增加外置低温泵数量、研发大抽速及大容量低温泵、提高加料系统可靠性、增加偏滤器超声分子束注入系统、提高离子回旋射频放电功率>50 kW等。通过这些改造,增加了泵的抽气能力及加料系统的加料能力,提高了器壁的清洗效果,使等离子体放电真空室压力达到约1.5×10^(-6) Pa。真空系统的成功升级,促进了EAST装置,如>400 s高约束模(H模)、1056s长脉冲等离子体等多项创世界纪录成果的获得,为未来ITER及聚变堆真空系统的设计与运行提供了重要的参考。

HL-2A装置硅化镀膜壁处理控制杂质行为研究

硅化壁处理能有效控制等离子体杂质和再循环水平,是HL-2A装置的常规壁处理手段。介绍了HL-2A装置的硅化壁处理系统以及硅化镀膜壁处理控制杂质行为的主要研究进展。HL-2A装置硅化壁处理系统包括离线硅化和在线硅化壁处理系统,分别采用辉光放电辅助沉积以及等离子体放电辅助沉积进行硅化镀膜壁处理。实验结果表明,离线硅化壁处理后氧杂质明显降低,但随着实验进行会逐渐增加,硅化镀膜寿命约150炮。为进一步开展研究,HL-2A采用了在线硅化壁处理技术,研究发现在线硅化壁处理过程中碳、氧、铁等杂质均有所减少。两种硅化方式下真空室内的H_(2)O、CO含量均明显降低,在线硅化壁处理耗气量远少于离线硅化。通过硅化壁处理,有效地提升了HL-2A等离子体性能,也为未来聚变装置稳态运行器壁的处理提供了重要技术参考。

HT-7超导托卡马克装置硼化实验研究

描述了一种无毒无爆炸的固态硼 (C2 B10 H12 )在HT 7超导托卡马克装置上进行的硼化实验 ,结果表明 :杂质成分显著降低 ,CO ,H2 O及碳化合物得到了有效抑制 ,Zeff从 4～ 5下降到约 1 5,杂质及工作气体再循环也明显减少 ,硼化为LHCD辅助加热提供了良好的壁条件。实验中对发现的问题及现象进行了探讨 ,提出了改进和完善的办法。

HL-1M的最新结果及等离子体性能的改善(英文)

本文综述了在ＨＬ－１Ｍ装置上进行的实验研究。完成了诸如壁处理、等离子体平衡控制、新加料技术、高功率辅助加热、改善约束等物理实验。把壁处理（硼化、硅化和涂锂）、偏压电极、弹丸注入、ＭＢＩ和ＬＨＣＤ等新技术引入后，获得了高品质、稳定、可重复的等离子体。最长的等离子体放电可持续４ｓ，且具有近似为矩形的放电波形。

氖杂质注入条件下CFETR芯部钨杂质浓度的模拟研究

由于低质量数材料不可接受的高腐蚀率以及氚共沉积的问题,未来聚变堆中更希望使用全钨壁。由于钨在芯部的高辐射冷却率,芯部的钨杂质浓度需要限制在非常低的水平(约10^(-5))。中国聚变工程试验堆(China Fusion Engineering Test Reactor,CFETR)要求其高功率稳态运行,全钨壁是优先考虑的方案。为了估计全钨壁CFETR的芯部钨杂质浓度,用边界等离子体物理模拟软件SOLPS(Scrape-off Layer Plasma Simulation)对下单零偏滤器位形不同氖气(Ne)辐射杂质注入速率下模拟得到边界等离子体背景,再利用蒙特卡罗杂质输运程序DIVIMP(DIVertor and IMPurity)对钨杂质的输运进行了模拟。当Ne注入速率较低、靶板温度仍然较高时,即使仅考虑靶板为钨材料,芯部钨杂质浓度依然过高。当外靶板峰值温度降低至约10 eV时,钨靶板对芯部钨杂质浓度的贡献降至可接受的水平;但当包含主等离子体室壁的贡献时,芯部钨杂质浓度仍然达到10^(-4)的水平。因此当Ne杂质注入速率较高时,过高的芯部钨杂质浓度主要来源于主等离子体室壁。未来的工作中需要进一步关注钨壁对芯部钨杂质浓度的影响。