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.     

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。     

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.     

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。     

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.     

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.     

Reduction of heat flux on divertor plates by remote radiative cooling in doublet III

Using a single null divertor configuration, heat flux intensity and its profile on the divertor plates as a function of plasma current and density were measured with an infrared camera and thermocouples. The vertical width of the heat flux on the divertor plates 2λ is ≈ 10 cm at the lower separatrix and is ≈ 5.5 cm at the upper separatrix. A diffusion coefficient D which is obtained from the measurement of the diffusion length across the scrape-off field lines is roughly proportional to and its magnitude is on the order of Bohm diffusion. The heat flux on the plates decreases by more than a factor of 5 with increasing electron density in the main plasma and is much smaller than that on the limiters in non-diverted plasmas. Only 3% of ohmic input power goes into the divertor plates at high density of the main plasma, while ≈ 20% goes in at low density. The decrease of heat flux is in good agreement with the increase of radiation loss in the divertor region. The heat flux on the divertor plates can be reduced by remote radiative cooling in high density discharges.     

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.     

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。     

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.     

Study of Divertor Heat Patterns Induced by LHCD L-Mode Plasmas Using an Infra-Red Camera System on EAST

Divertor heat patterns induced by Lower Hybrid Current Drive(LHCD) L-mode plasmas are investigated using an infra-red(IR) camera system on an Experimental Advanced Superconducting Tokamak(EAST). A two-dimensional finite element analysis code DFlux is used to compute heat flux along the poloidal divertor target and corresponding quantities. Outside the Origin Strike Zone(OSZ), a Second Peak Heat Flux(SPHF) zone, where the heat flux is even stronger than that at the OSZ, appears on the lower-outer(LO) divertor plates with LHCD and disappears immediately after switching off the LHCD. The main heat-flux shifts from the SPHF zone towards the OSZ when the divertor configuration converts from double null to lower single null, indicating that the growth of the SPHF zone is apparently affected by a plasma magnetic configuration. The heat patterns on the LO divertor plates are observed to be different from that on the lower-inner(LI) targets as the SPHF zone appears only on the LO divertor target. It is also found that the heat flux at the SPHF zone was obviously enhanced after the Supersonic Molecule Beam Injection(SMBI) pulse.     

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.     

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.     

The in-vessel components of the experiment WENDELSTEIN 7-X

The in-vessel components of the WENDELSTEIN 7-X stellarator consist of the divertor components and the wall protection with its internal cooling supply. The main components of the open divertor are the vertical and horizontal target plates which form the pumping gap, the cryo-vacuum pumps and the control coils. The divertor volume is closed by graphite shielded baffle modules and with divertor closures. All these components are designed to be actively water-cooled. For the first commissioning phase planned in 2014, an inertial-cooled test divertor will be installed instead of the actively water-cooled high heat flux divertor. The wall protection consists of graphite-protected heat shields in the higher loaded areas and stainless steel panels in the lower loaded regions. The wall protection cooling circuits are connected through 80 supply-ports via so-called “plug-ins”. It is envisaged to protect the diagnostic ports by panel-type port-liners. Special graphite-shielded port liners are used on the diagnostic injector and the neutral beam injector ports. The in-vessel components are mainly manufactured and tested at the Max-Planck-Institute für Plasmaphysik in its Garching workshop. Panels, high heat flux target elements and control coils are delivered by industrial partners. Manufacturing of the KiP (“Komponenten im Plasmagefäß”) is in plan. Delivery of the components will be in time.     

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.     

Simulation of Small Size Divertor Tokamak Plasma Edge Including Self-Consistent Electric Fields

The B2.SOLPES.0.5.2D code (Braams, Contrib Plasma Phys 36:276, 1996; Rozhansky and Tendler, Rev Plasma Phys 19:147, 1996) is applied for modeling SOL (Scrape off Layer) plasma in the small size divertor tokamak. Detailed distributions of the plasma heat flux and other plasma parameters in SOL, especially at the target plate of the divertor are found by modeling. The modeling results show that most of the electron heat flux and small part of ion heat flux arrive at target plate of the divertor, while, a large part of the ion heat flux and part of electron heat flux arrive at the outer wall. Also analysis of the role of poloidal E × B drifts in the redistribution of edge plasma is fulfilled.     

Characterization of transient particle loads during lithium experiments on the National Spherical Torus Experiment

Transient events such as Edge Localized Modes (ELMs) or disruptions can lead to large particle and power loads on the divertor plates of tokamak experiments. These events can cause significant erosion and are detrimental to the lifetime of the plasma facing components. Understanding the impact of ELMs remains a complex problem and a major challenge. In this study, an effort is made to characterize these ELMs and other transients based on their characteristics using a particle flux probe and surface temperature measurements from a dual-band IR camera. Typically, the temporal evolution of an ELM from the particle flux probe is characterized by a steep rise and a gradual decrease of current signal. This burst like structure is seen by the Langmuir probes as a rise in the ion saturation current with a width of a few milliseconds. This study entails gathering statistics of typical ELM-like events for various shots in order to assess the typical loading of ELMs on the Liquid Lithium Divertor (LLD) that was installed in the FY10 run campaign. Later, the power deposition profiles during ELMs are also characterized from IR camera measurements for certain discharges to find that only 15% of the energy flux arrives at the divertor target before the surface temperature reached its maximum value. Finally, a correlation was found between the particle flux from the probes during the ELMs and the neutral particle flux from D_α signal indicating the utility of the particle flux probe as a means to characterize ELMs.     

Super X divertors for solving heat and neutron flux problems of fusion devices

We present a new magnetic geometry, called the Super X divertor (SXD), that could potentially solve the enormous heat exhaust problem of next-generation high power-density experiments and fusion reactors. With only small changes in net coil currents, the axisymmetric SXD modification of the standard divertor (SD) coils greatly increases the divertor radius, the line length, and the plasma-wetted area. The lower B at large R decreases parallel heat flux and hence lowers the plasma temperature at SXD plates to below 10 eV, allowing higher divertor radiation fractions. The SXD could safely exhaust five times more heat than an SD, is unique in allowing adequate shielding of divertor target from neutron damage, and can enable much improved, reactor-relevant core plasma performance.