Design and installation of the closed helical divertor in LHD

Neutral particle behavior in the Large Helical Device heliotron has been investigated to conduct the effective particle control using the intrinsic helical divertor. It was revealed that the torus in-out asymmetry observed in the neutral pressure distribution depended on the divertor particle flux distribution, and thus, on the operational magnetic configuration. It was also revealed that the neutral pressure in the vacuum vessel in LHD was below 0.1 Pa. Degradation of the plasma confinement with increasing of the neutral pressure was observed, and that suggested the effective particle control is necessary for the sustaining of long discharge with high performance plasma and the further improvement of the confinement. The modification of the open helical divertor to the closed one was investigated for the particle control using helical divertor by using EIRENE code. Results of the calculation showed that proper rearrangement of divertor plates and additional components, such as dome structure make the neutral particles to be compressed well in the divertor region, and effective divertor pumping to be possible. Based on the simulation and experimental results, design of the closed helical divertor was completed and it will be partially installed in the Large Helical Device before the experimental campaign in 2010.     

Formation of Edge Transport Barriers by L-H Transition and Large Reversed Plasma Current on LHD

On the Large Helical Device （LHD） where nested magnetic surfaces are surrounded by the ergodic field layer, edge transport barrier （ETB） was produced in neutral-beam-injection （NBI） heated plasmas through transition and non-transition processes. The former case is the ETB formation by L-Htransition, where characteristics of L-H transition observed in a tokamak plasma are clearly recognized. The confinement improvement is the modest （- 10%）, compared with the ISS95 international stellarator scaling. The threshold power for the transition is comparable or slightly lower than the ITER scaling law established by tokamaks and compact tori. The ETB is formed inside the ergodic field layer of the vacuum field. The ETB formation destabilizes edge coherent modes such as m/n = 1/1, 2/3 and 1/2, of which rational surfaces are in the magnetic hill. The formed ETB is partially and transiently destroyed by these coherent edge MHD modes and edge localized modes （ELMs） typically observed in Ha signals. The latter ETB is observed in a plasma with large reversed NBI-driven current more than 100 kA at Bt = 1 T. In these plasmas, the edge magnetic shear is enhanced by the current and the rotational transform in the core region is expected to be appreciably reduced. Thus reduced rotational transform in the plasma central region will enhance outward heat and particle fluxes toward ergodic edge layer. The ETB with steep electron temperature gradient up to - 5 keV/m is formed by blocking enhanced outward heat flux.     

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.     

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.     

Divertor detachment operation in helium plasmas with ITER-like tungsten divertor in EAST

Detachment in helium (He) discharges has been achieved in the EAST superconducting tokamak equipped with an ITER-like tungsten divertor. This paper presents the experimental observations of divertor detachment achieved by increasing the plasma density in He discharges. During density ramp-up, the particle flux shows a clear rollover, while the electron temperature around the outer strike point is decreasing simultaneously. The divertor detachment also exhibits a significant difference from that observed in comparable deuterium (D) discharges. The density threshold of detachment in the He plasma is higher than that in the D plasma for the same heating power, and increases with the heating power. Moreover, detachment assisted with neon (Ne) seeding was also performed in L- and H-mode plasmas, pointing to the direction for reducing the density threshold of detachment in He operation. However, excessive Ne seeding causes confinement degradation during the divertor detachment phase. The precise feedback control of impurity seeding will be performed in EAST to improve the compatibility of core plasma performance with divertor detachment for future high heating power operations.     

Impurity Transport in a Simulated Gas Target Divertor

Future generation fusion reactors and tokamaks will require dissipative divertors to handle the high particle and heat loads leaving the core plasma (100–400 MW/m² in ITER). A radiative divertor is proposed as a possible scenario, utilizing a hydrogen target gas to disperse the plasma momentum and trace impurity radiation to dissipate the plasma heat flux. Introducing an impurity into the target hydrogen gas enhances the radiative power loss but may lead to a significant impurity backflow to the main plasma. Thus, impurity flow control represents a crucial design concern. Such impurity flows are studied experimentally in this thesis. The PISCES-A linear plasma device (n 3 × 10¹⁹ m^–3, kT _e 20 eV) has been used to simulate a gas target divertor. To study the transport of impurities, a trace amount of impurity gas (i.e., neon and argon) is puffed near the target plate along with the hydrogen gas. Varying the hydrogen gas puffing rate permits us to study the effects of various background plasma conditions on the transport of impurities. A 1-1/2-D fluid code has been developed to solve the continuity and momentum equations for a neutral and singly ionized impurity in a hydrogen background plasma. The results indicate an axial reduction in the impurity concentration upstream from the impurity puffing source. Impurity entrainment is more effective for higher hydrogen target pressures (and for higher hydrogen plasma densities). However, if there is a reversal of the background plasma flow, impurity particles can propagate past the plasma flow reversal point and are then no longer entrained.     

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.     

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.     

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.     

Influence of the X-point location on edge plasma transport in the J-TEXT tokamak with a high-field-side single-null divertor

High-density experiments in the high-field-side mid-plane single-null divertor configuration have been performed for the first time on J-TEXT.The experiments show an increase in the highest central channel line-averaged density from 2.73 x 1019 m-3 to 6.49 x 1019 m-3,while the X-point moves away from the target by increasing the divertor coil current.The corresponding Greenwald fraction rises from 0.50 to 0.79.For the impurity transport,the density normalized radiation intensity(absolute extreme ultraviolet and soft x-ray)of the central channel density decreased significantly(＞50％)with an increase in the plasma density.To better understand the underlying physics mechanisms,the 3D edge Monte Carlo code coupled with EIRENE(EMC3-EIRENE)has been implemented for the first time on J-TEXT.The simulation results show good agreement with the experimental findings.As the X-point moves away from the target,the divertor power decay length drops and the scrape-off layer impurity screening effect is enhanced.     

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.     

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.     

Measurement of the Electron Bernstein Wave Emission with One of the Power Transmission Lines for ECH in LHD

Possibility of the measurement of radiated waves derived from the thermally emitted electron Bernstein wave （EBW） is numerically investigated based on the assum...     

Magnetic Islands Observed by a Fast-Framing Tangentially Viewing Soft X-Ray Camera on LHD and TEXTOR

The formation of magnetic islands within plasmas confined magnetically within the tori has significant influence upon their confinement and stability. To obtain an experimental insight into the formation and dynamics of such island structures we employed a fast framing camera viewing the plasma tangentially in the toroidal direction. The toroidal viewing direction gives the advantage in that the islands are viewed almost tangentially and this greatly facilitates the reconstruction of the local data from the line integrated ones. We discuss an effective method to do inversion. To study the fluctuations seen in the video images we perform a singular value decomposition, and then we use a truncated least square method to infer their pictures in space.     

Functions of intrinsic magnetic field line structures on lithium ion transport in the LHD plasma periphery

Plasma discharge operation with lithium coating suggests that the lithium effectively control neutral particles in the plasma periphery, which can lead to improvement of plasma parameters. The effect of lithium coating on the large helical device (LHD) for a closed helical divertor configuration is discussed from viewpoints of neutral particle and impurity ion transport in the plasma periphery. It shows that the closed helical divertor configuration can enhance the neutral particle density in the divertor region, which is enough to achieve efficient particle control, and that it can effectively confine neutral lithium atoms near divertor plates. A one-dimensional impurity (lithium) ion transport analysis along magnetic field lines on divertor legs indicates that the friction force due to the plasma flow from the main plasma is dominant over the thermal force caused by the temperature gradient on the divertor legs, which prevents lithium ion contamination in the main plasma and excessive cooling of the plasma temperature in an ergodic layer. The analysis shows that the lithium coating is compatible with LHD plasma discharge operation for the closed helical divertor configuration.     

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。     

Spectroscopic Studies on Impurity Transport of Core and Edge Plasmas in LHD

Spectroscopic diagnostics have been extensively developed for studies of impurity and neutral particle transports at core and edge plasmas in LHD. Diagnostics of core plasmas are similar to a tokamak case, i.e., Zeff from visible bremsstrahlung, K-x-ray measurements from xray spectroscopy using Si（Li） detectors and a compact crystal spectrometer, and high-Z impurity diagnostics from VUV spectroscopy using a flat-field EUV spectrometer. A combination of impurity pellet injection and visible bremsstrahlung is an active tool for determination of the diffusion coeffici＇ent D and convective velocity V. Using this tool the spatial structures of D and V are obtained and discussed with a neoclassical effect. On the other hand, the spectroscopic method for edge diagnostics is considerably different from the tokamak case because of the existence of a thick ergodic layer in addition to the z-points necessarily included into the diagnostic chord view. In order to break this negative situation, Zeeman and polarization spectroscopy are adopted to LHD edge plasmas. As a result, 2-dimensional emission contours of HeI and Ha are successfully obtained. Laser absorption spectroscopy is tried to measure hydrogen neutrals directly. Radial profiles of edge impurities are also measured with a mirror-assembled 3 m VUV spectrometer. Recent results of and progress in LHD spectroscopy are briefly reviewed.     

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.     

Investigation of turbulence rotation in the SOL and plasma edge of W7-X for different magnetic configurations

The W7-X stellarator is optimized with respect to neoclassical transport. Therefore turbulent transport plays an important role. It is equipped with an inertial cooled graphite divertor which intersects the island chain at the plasma edge depending on the magnetic configuration. Additional control coils and the plasma current modify the iota profile at the plasma edge and shift the position of the island chain. To monitor the effects on the poloidal propagation velocity in the scrape-off layer(SOL) and the plasma edge, an O-mode Poloidal Correlation Reflectometer(PCR) is used which simultaneously monitors the propagation of low-k turbulence. Operating in the density range of 0.6?×?10~(19) m~(-3) to2?×?10~(19) m~(-3) it covers a large part of the SOL and the plasma edge and allows for the experimental determination in the last closed flux surface(LCFS) and the associated shear layer in low to middensity discharges. In this paper it is shown that the propagation in the shear layer and its vicinity is determined best, when based on an elliptical model. Different magnetic configurations with magnetic edge topology of five independent islands for ι?=?1 and six linked islands for ι?=?0.81 are investigated. Also the effects of the plasma current and additional control coils on the edge magnetic topology are studied. The coherence spectra of antenna pairs for different poloidal separations is investigated. Using a decomposition method for the measured coherence spectra the characterization of turbulence spectra is possible with respect to e.g. broad band turbulence and quasi coherent modes.A strong reduction of the broad band turbulence is observed in the vicinity of the LCFS which is evidence for the suppression of low-k turbulence at the shear layer.