Preliminary Design of GDC System for KTX Device

KeDa Torus for eXperiment (KTX) is a reversed field pinch magnetic confinement fusion research device whose main parameters are between in the RFX and MST. The base vacuum of KTX is 1 × 10^?6 Pa. Six sets of turbomolecular pumps parallel installed in the six horizontal ports which served as the main pumping system for KTX and the diameters of the ports are 0.15 m. Before plasma discharge, glow discharge cleaning (GDC) system is applied to clean C, O and hydrocarbon impurity on the vacuum vessel (VV) surfaces of KTX. An inflation system and residual gas analyzer system are designed to supply the working gas and monitoring the effect of GDC respectively. According to the GDC experiment practice, the working gas pressure of the KTX GDC system is designed as 0.3 Pa, with average current density of 0.15 A/m². Two sets of the GDC probes are installed in KTX horizontal ports symmetrically with interval angle of 180º and the input current of each anode is 1.6 A. According to the current density distribution, the centre of the VV cross section is the superior working area for GDC anode, a screw-nut pairs with the cooperation of bellows can transfer the anode from its storage position to its working position, and the stroke of the screw-nut pairs is 0.5 m. Based on the temperature rise calculation, the maximum equilibrium temperature of the anode during glow discharge is about 275 °C (under 200 °C baking). The thermal stresses caused by the temperature distribution on the anode’s components especially in the vacuum brazing areas are inspected during GDC process. All the simulation results show that the structure and base material of the KTX GDC anode can work normally without additional active cooling system.     

Magnetic configuration flexibility of snowflake divertor for HL-2M

HL-2M (Li, 2013 [1]) is a tokamak device that is under construction. Based on the magnetic coils design of HL-2M, four kinds of divertor configurations are calculated by CORSICA code (Pearlstein et al., 2001 [2]) with the same main plasma parameters, which are standard divertor, exact snowflake divertor, snowflake-plus divertor and snowflake-minus divertor configurations. The potential properties of these divertors are analyzed and presented in this paper: low poloidal field area around X-point, connection length from outside mid-plane to the primary X-point, target plate design and magnetic field shear. The results show that the snowflake configurations not only can reduce the heat load at divertor target plates, but also may improve the magneto-hydrodynamic stability by stronger magnetic shear at the edge. A new divertor configuration, named “tripod divertor”, is designed by adjusting the positions of the two X-points according to plasma parameters and magnetic coils current of HL-2M.     

A Study on Helium Glow Discharge Cleaning in the HL-1M Tokamak

Based on the principle of ion-bombarded reemission and sputtering desorption.the GLow Discharge Cleaning with helium (GDC(He)is an effective method for controlling the recycle of H on the chamber wall,Carbon(C),Oxygen(O)impurity and improving the wall conditioning in HL-1M tokamak,It is characterized by simplicity without magnet and safety,compared,with Taylor Discharge Cleaning (TDC) ,Altenating Current glow discharge Cleaning (AC) ,Electron Cyclotron Resonance-Deischarge Cleaning (ECR-DC),Compared with bake-out degassing ,the wall has a higher degassing rate during GDC(He) and a lower impurity concentration in vacuum chambers after GDC(He) .Cleaning Patterns have been developed dominantly for de-oxdization ,decarbonization and de-hydrogenization,The cleaming parameters for H recycle on the wall are also presented,This paper mainly describes the GDC system along with its parameters,breakdown voltage,volt-ampere characteristic,the range of operation safe and suitable cleaning pattens in the HL-1M tokamak finaly concluding with some suggestion on HL-2A GDC.     

HL-2M Divertor Geometry Exploration with SOLPS5.0

One of the critical issues to be solved for HL-2M is the power and particle exhaust. Divertor target plate geometry strongly influences the plasma profiles by controlling the neutral recycling pattern, which has in turn a strong effect on the symmetry and stability of the divertor plasma and finally on the whole edge region. The numerical simulation software SOLPS5.0 Pack- age is used to design and explore the divertor target plates for HL-2M. We choose two divertor geometries, and assess the heat flux on the target plates and first wall, then further discuss the di- vertor plasma parameters, and how private flux baffling affects both neutral recirculation pattern and pumping efficiency.     

HL-2A托卡马克装置的工程和实验概况

中国环流器二号A装置(HL-2A)是核工业西南物理研究院2002年投入实验运行的托卡马克,它是我国第1个具有偏滤器、等离子体截面具有一定垂直拉长的托卡马克.HL-2A的磁体使用铜导体,具有良好的灵活性和等离子体的可近性,其极向场线圈全部位于环向场线圈之内,位于真空室内的偏滤器的成形线圈可建立双零和单零的偏滤器位形.HL-2A已发展了30多套先进的等离子体诊断系统和总功率4 MW的辅助加热系统,加料技术得到持续发展.随着上述系统的建设和放电综合控制技术的提高,HL-2A装置已获得了高约束模式,这为开展先进托卡马克(AT)物理实验,ITER和聚变堆的科学、技术和工程问题等的研究奠定了基础.HL-2A也成为国际上最活跃的中型托卡马克,为国际托卡马克物理活动(ITPA)作出了积极贡献.     

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。     

中国环流器二号A(HL-2A)装置物理设计

文章是关于中国环流器二号A(HL-2A)装置物理设计的总结报告,包括以下几方面的内容:分析计算等离子体截面变形及由截面拉长引起的垂直不稳定性,提出对HL-2A极向磁场线圈电流和控制系统的要求;研究通过中性束注入加热(NBI)和低混杂波电流驱动(LHCD)实现等离子体剖面控制,模拟并设计HL-2A的高性能的运行模式;分析HL-2A先进约束位形(RS位形)下的磁流体力学不稳定性,为实现高性能模式稳态运行的等离子体控制指出方向;同时,利用数值模拟分析HL-2A偏滤器等离子体性能,为偏滤器的改进提供依据。     

Design and test of the KSTAR vacuum pumping system

The Korea Superconductor Tokamak Advanced Research (KSTAR) device is a tokamak mainly composed of a vacuum vessel, superconducting magnets, and cryostat. The internal volume of the vacuum vessel is about 110 m³ with a target pressure of 1 × 10⁻⁶ Pa, while the volume of the cryostat is 450 m³ with a target pressure of 5 × 10⁻³ Pa. To attain these target pressures, two identical vacuum pumping systems consisting of dry pumps, mechanical booster pumps, turbo-molecular pumps, and cryopumps were installed. The control system of the vacuum pumping systems was built using the experimental physics and industrial control system (EPICS), which has various merits such as easy access, convenient extension and flexible integration. The pump-down test of the pumping ducts was successfully executed under the control of the EPICS system.     

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.     

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.     

Upgrade of an integrated Langmuir probe system on the closed divertor target plates in the HL-2A tokamak

A newly designed divertor Langmuir probe diagnostic system has been installed in a rare closed divertor of the HL-2A tokamak and steadily operated for the study of divertor physics involved edge-localized mode mitigation, detachment and redistribution of heat flux, etc. Two sets of probe arrays including 274 probe tips were placed at two ports (approximately 180° separated toroidally), and the spatial and temporal resolutions of this measurement system could reach 6 mm and 1 μs, respectively. A novel design of the ceramic isolation ring can ensure reliable electrical insulation property between the graphite tip and the copper substrate plate where plasma impurities and the dust are deposited into the gaps for a long experimental time. Meanwhile, the condition monitoring and mode conversion between single and triple probe of the probe system could be conveniently implemented via a remote-control station. The preliminary experimental result shows that the divertor Langmuir probe system is capable of measuring the high spatiotemporal parameters involved the plasma density, electron temperature, particle flux as well as heat flux during the ELMy H-mode discharges.     

Preliminary concept design of the divertor remote handling system for DEMO power plant

This paper is based on the remote maintenance system project (WPRM) for the demonstration fusion power reactor (DEMO). Following ITER, DEMO aims to confirm the capability of generating several hundred of MW of net electricity by 2050. The main objective of these activities is to develop an efficient and reliable remote handling (RH) system for replacing the divertor cassettes.This paper presents the preliminary results of the concept design of the divertor RH system. The proposed divertor mover is a hydraulic telescopic boom driven from the transportation cask through the maintenance tunnel of the reactor. The boom is divided in three sections of 4 m each, and it is driving an end-effector in order to perform the scheduled operations of maintenance inside the vacuum vessel.Two alternative design of the end effector to grip and manipulate the divertor cassette are also presented in this work. Both the concepts are hydraulically actuated, basing on the ITER previous studies. The divertor cassette end-effector consists of a lifting arm linked to the divertor mover, a tilting plate, a cantilever arm and a hook-plate.The main objective of this paper is to illustrate the feasibility of DEMO divertor remote maintenance operations.     

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.     

HL-2A Tokamak Edge Modeling with B2

The outer divertor plasma of HL-2A and its associated scrape-off plasma have been simulated using a two-dimensional multi-species fluid code of Braams with a simplified neutral gas model. HL-2A has a double-null closed divertor in separate divertor chambers above and below the nearly circular plasma tours. The computed numerical grid is developed according to an ideal magnetic surface. The calculation is involved only with pure hydrogen plasma. The emphasis has been focused on parametric studies involving variation of the assumptions made for the core plasma. The peak temperatures and the heat flux near the target are of the principal concern。     

Design of the stripping unit and the electromagnetic analysis unit for the E//B NPA on HL-2A/2M tokamak

An E//B neutral particle analyzer is under development for fast ion diagnosis on HL-2A/2 M tokamak. The stripping unit is composed of a stripping room (equipped with two differential tubes and a gas supply bellows), a vacuum chamber and a vacuum pumping system. The stripping efficiency of the stripping room is calculated in the form of global efficiency R × f₊₁, where R is the non-scattered-away rate and f₊₁ is the fraction of charge state +1. The magnetic field of the E//B analyzer is produced with a permanent magnet. The yoke and the poles of the magnet are made of mild steel and the magnet plates are made of NdFeB. The magnetic poles are specially designed to focus the ion trajectories and to increase the use rate of the magnet. The shape of the magnet and the electric plates are carefully designed so that the ions are dispersed into two lines of H⁺ and D⁺ on the detector plane. For each line, the energy increases from 10 to 200 keV from one side to another.     

CYCIAE-100紧凑型回旋加速器低温板排气系统理论和实验研究

根据CYCIAE-100紧凑型回旋加速器的结构特点,设计了一套大抽速低温板排气系统,插入到CYCIAE-100紧凑型回旋加速器主磁铁谷区中,该套排气系统将使CYCIAE-100紧凑型回旋加速器加速腔内真空度好于5×10^-6 Pa。目前该套低温板排气系统已设计完成,用Monte-Carlo方法对其进行了优化,并进行了加工、安装和初步调试,调试时低温冷板上温度达19 K,CYCIAE-100紧凑型回旋加速器加速腔内真空度达到8×10^-6 Pa。     

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.     

Thermal-hydraulic analysis of the HL-2M divertor using an homogeneous equilibrium model

The heat flux of the HL-2M divertor would reach 10 MWm-2 or more at the local area when the device operates at high parameters.Subcooled boiling could occur at high thermal load,which would be simulated based on the homogeneous equilibrium model.The results show that the current design of the HL-2M divertor could withstand the local heat flux 10 MW m-2 at a plasma pulse duration of 5 s,inlet coolant pressure of 1.5 MPa and flow velocity of 4 m s-1.The pulse duration that the HL-2M divertor could withstand is closely related to the coolant velocity.In addition,at the time of 2 min after plasma discharge,the flow velocity decreased from 4 m s-1 to 1 m s-1,and the divertor could also be cooled to the initial temperature before the next plasma discharge commences.     

100 MeV强流质子回旋加速器超高真空系统研制

中国原子能科学研究院建成了一台强流质子回旋加速器，其引出能量为100 MeV，流强为200 μA。为减小粒子加速时束流损失的目的，其粒子加速腔内工作真空度要求为6.7×10^-6 Pa。由于是紧凑型加速器结构，该加速器能提供给真空系统利用的通路有限，为此主真空系统设计为内置式低温冷板结合商业低温泵的排气方案以增加系统整体的抽气能力。设计、加工完成的真空系统已成功应用于100 MeV强流质子回旋加速器上，为加速器的束流调试和正常供束提供了有利的保障。     

Recent progress of NSTX lithium program and opportunities for magnetic fusion research

Lithium wall coating techniques have been experimentally explored on National Spherical Torus Experiment (NSTX) for the last six years. The lithium experimentation on NSTX started with a few milligrams of lithium injected into the plasma as pellets and it has evolved to a dual lithium evaporation system which can evaporate up to ～160 g of lithium onto the lower divertor plates between re-loadings. The unique feature of the NSTX lithium research program is that it can investigate the effects of lithium coated plasma-facing components in H-mode divertor plasmas. This lithium evaporation system has produced many intriguing and potentially important results. In 2010, the NSTX lithium program has focused on the effects of liquid lithium divertor (LLD) surfaces including the divertor heat load, deuterium pumping, impurity control, electron thermal confinement, H-mode pedestal physics, and enhanced plasma performance. To fill the LLD with lithium, 1300 g of lithium was evaporated into the NSTX vacuum vessel during the 2010 operations. The routine use of lithium in 2010 has significantly improved the plasma shot availability resulting in a record number of plasma shots in any given year. In this paper, as a follow-on paper from the 1st lithium symposium [1], we review the recent progress toward developing fundamental understanding of the NSTX lithium experimental observations as well as the opportunities and associated R&D required for use of lithium in future magnetic fusion facilities including ITER.