对JT—60U的紧凑、半闭合W型偏滤器系统的结构计算

在JT－60U上已设计、制造并安装一种紧凑、半闭合W型偏滤器系统，以替代开式偏滤器系统。这种新的系统由倾斜的偏滤器、圆顶和缓冲板组成。为了满足结构上的要求，采用一种灵活的电绝缘气体密封分段式结构。用FEM编码，通过电磁和结构分析证实了等离子体破裂时系统结构的完整性。上述分析考虑了晕电流。在偏滤器系统中由于气体密封结构采用电绝缘致使感应电磁力显著下降。另外，因为系统是分段式结构，发现每个部件单位上的感应电磁力是有限的。获得的最大应力强度及其范围在允许值内，烘烤运行期间偏滤器系统和真空室之间的温差所引起的热应力也是令人满意的。而且，热和热应力分析表明，面向等离子体部件有足够的结构完整性。     

FEB实验混合堆偏滤器工程结构与热工分析

在FEB偏滤器物理研究的基础上,描述FEB偏滤器的结构设计与热工分析。在工程概要设计(FEB-E)阶段,偏滤器的结构从开式固板靶优化为闭式气室靶,以改善偏滤器的杂质控制和增强原子损失过程。偏滤器运行在喷气和注杂组合下的脱靶等离子体或部分脱靶等离子体模式(Partial Detached Plasma Mode)。应用改进了的NEWT1D编码模拟了喷气和注杂状态下删削层/偏滤器中等离子体及杂质的输运。着眼于杂质居留和杂质辐射,优化了喷气口的位置。偏滤器主体结构由48个气室模件环绕组成。应用COSMOS/M-HSTAR编码对偏滤器靶板进行了热工计算,结果表明以4MPa压力的氦气对靶板进行径向循环冷却是可行的。     

FEB混合堆偏滤器的物理研究

描述FEB偏滤器的物理设计研究。在工程概要设计(FEB-E)阶段,偏滤器的结构从开式优化为闭式,以改善偏滤器的杂质控制和增强原子损失过程。偏滤器运行在喷气和注杂组合下的脱靶等离子体或部分脱靶等离子体模式(Partial Detached Plasma Mode)。硼被选为注入杂质。应用改进了的NEWT1D编码模拟了喷气和注杂下删削层/偏滤器中等离子体及杂质的输运。着眼于杂质居留和杂质辐射,优化了喷气口的位置。在辐射偏滤器概念下,应用两点输运编码和杂质辐射模型估算了等离子体压力。计算表明:偏滤器靶前等离子体压降系数f_p不仅和辐射份额f_(rad)有关,而且和删削层驻点密度n_s密切相关。     

CFETR偏滤器概念设计

偏滤器是托卡马克装置——中国聚变工程实验堆(China Fusion Engineering Test Reactor,CFETR)的核心部件之一,是高温等离子体与材料直接接触的过渡区域。本文在满足CFETR总体设计要求和设计依据的前提下,针对不同的等离子体位型(类ITER和雪花等离子体位型),设计了同一盒体支撑的3种偏滤器结构,即类ITER偏滤器、雪花偏滤器和类ITER-雪花偏滤器,并对类ITER偏滤器结构进行了详细设计,包括第一壁设计、两侧支撑轨道设计、冷却系统设计和抽气系统设计。通过计算磁力线膨胀系数,验证了3种偏滤器结构设计的合理性,为偏滤器结构进一步优化和改进提供了必要的理论依据。     

ASDEX—U和JET上偏滤器几何形状对边界及芯部等离子体性能的影响

将来象ＩＴＥＲ之类的环形磁聚变装置的一个关键因素是偏滤器的设计,这种偏滤器能在维持好整体等离子体性能的情况下安全排出粒子和能量。相应的、优化的偏滤器的确定和设计是正进行的国际ＩＴＥＲ研究和发展工作的一项主要任务。为了给偏滤器优化提供一个深厚的物理基础,一些不同的偏滤器几何形状正在一些主要的托卡马克装置上试验。本文描述了ＡＳＤＥＸ－Ｕ和ＪＥＴ上这些偏滤器改进对等离子体性能的影响,结论是,增强封闭性可     

H模式的主动控制

本文讨论报有关Ｈ模式的主动控制研究现状，其中包括稳态Ｈ模式运行的密度控制，为降低偏滤器势负载的热通量控制和为进一步改善约束和ＭＨＤ稳定性的输运任务的控制。ＥＬＭ活性的控制对Ｈ模式中的密度控制极其重要本文讨论了控制ＥＬＭ的一些例子。主等离子体和偏滤器等离子体的远距离辐射冷却是降低偏滤器热负载必要的。     

JET运行的工程经验

ＪＥＴ初期设计的特有灵活性在于允许工程升级和改进，以研究许许多等离子体和聚变物理问题，最近，主要的改进就是配置了一套轴对称零抽气偏滤器（ＭａｒｋＩ）它成功地运行在１９９４年－１９９５年实验期间，根据偏滤器优化方案，一个新的，更为封闭的偏滤器现已安装完成（ＭａｒｋⅡ）这个偏滤器显示了良好的功率处理能力，并大大改善了中性粒子的滞留，这个新设计的一个关键特性是可能用按照扩展的Ｄ－Ｔ运行设计的全遥控技术替     

2×2棒束内超临界水传热特性数值研究

复杂通道是介于典型通道和典型栅元之间的几何结构,在通道形状、加热方式等方面更接近于原型组件。本文以小棒束2×2结构为分析对象,采用计算流体力学(CFD)工具分析组件内的温场特征,并讨论热流密度、质量流速等热工参数对通道平均换热系数的影响。结果表明,减小热流密度、增加质量流速都会增强换热。棒束结构的壁面温度分布存在明显的周向不均匀性,固体壁面导热会有效抑制这种周向不均匀性,进而增强通道的换热能力。     

窄缝通道内支撑柱对传热性能的影响

将计算流体动力学方法应用于窄缝通道内支撑柱对传热性能的影响分析。用CFX程序数值模拟有和无支撑柱两种情况下窄缝通道内的冷却剂流动，通过比较相同热流密度工况下加热壁面出口附近相应峰值点温度来分析支撑柱对通道内冷却剂冷却能力的影响。结果表明：支撑柱对该通道的临界热流密度的影响较小，可在实验中采用支撑柱以防止窄缝通道变形。     

HL-2M偏滤器热氮气烘烤仿真及优化

为提供高质量的等离子体真空运行环境,需对偏滤器进行高温烘烤。根据热传导与对流换热方程对偏滤器的烘烤过程进行了数值模拟及优化。结果表明:当热氮气等质量流量控制时,偏滤器回路压力损失逐渐增大,各部件烘烤温度爬升速率呈线性增加;当热氮气等体积流量控制时,偏滤器回路压力损失逐渐降低,各部件烘烤温度爬升呈线性增加。当初始条件近似相等(等质量流量为3×10~(-3)kg/s和等体积流量为4.8×10~(-4)m~3/s)时,前者的部件温升速率略低于后者,但各部件烘烤过程中最大温差均未超过90℃。     

Influence of multiple jet cooling on the heat transfer and thermal stresses in DEMO divertor cooling finger

The divertor concept for DEMO fusion reactor is based on modular design cooled by multiple impinging jets. Such divertor should be able to withstand a surface heat flux of at least 10 MW/m² at an acceptable pumping power. To reduce the thermal loads the plasma-facing side of the divertor is build up of numerous small cooling fingers. Each cooling finger is cooled by an array of jets blowing through the holes on the steel cartridge.The size, number and arrangement of jets on the cartridge influences the heat transfer and pressure drop characteristics of the divertor. Five different cartridge designs are analyzed in the paper. The most critical parameters, such as structure temperature, heat removal ability, pressure drop, cooling efficiency and thermal stress loadings in the cooling finger are predicted for each cartridge design. A combined computational fluid dynamics and structural model was used to perform the necessary numerical analyses. The results have shown that the cartridge design with the best heat transfer and pressure drop characteristics is not also the most favorable choice from the point of view of minimum stress peaks.     

The WEST project: PFC shaping solutions investigated for the ITER-like W divertor

Among major issues for PFCs design, the impact of leading edges (exposed surface) which would be directly intersected by particles following magnetic field lines at glancing incident angles in the high heat flux areas is much discussed. This paper presents the key outcome of a thermal analysis performed on different shaping solutions for the ITER-like W monoblocks occurred for the components of the WEST (W Environment for Steady state Tokamak) divertor which could shadow any direct leading edge and to counteract a potential misalignment due to assembly tolerance. The results, in terms of surface temperature rise and wall heat flux into the cooling channel, are discussed for magnetic field lines incident at glancing angles expected in the higher heat flux regions of divertor (i.e. close to the strike point regions) and for perpendicular incident heat flux up to 20 MW/m².     

Design and cooling of the edge segments of the DEMO divertor target plates

In support of shadowing of the divertor target plate edges in toroidal direction against damage caused by the incident particles, the fingers at the boundary of the target plate should ideally form a flat surface. The reference cooling fingers are of hexagonal shape and when assembled together, their edge boundary cannot be flat. Therefore, the boundary segments need to be designed in a different way. Three possible designs are investigated: non-symmetric pentagonal fingers and two square-shaped fingers of different sizes, all cooled by the same type of concentric cartridge as in the reference design. Their heat transfer performance is analyzed from the point of view of maximum allowable temperature of the thimble structure. The computational fluid dynamics (CFD) analysis is performed to obtain the minimum mass flow rate of the coolant which is necessary to keep the structure's temperature below the permissible limit at an acceptable pressure loss.     

Numerical Analysis of Fatigue Behavior of ITER-Like Monoblock Divertor Interlayer Under Coupled Heat Loads

High-confinement mode is a very prominent operation style for future fusion device due to its unique advantages. However, the conjuncted edge localized modes (ELMs) are very difficult to control so that divertor plates are very prone to suffer both stationary high heat flux (HHF) loads of long-pulse operating mode and transient shock loads of ELMs. Most previous researches focus on degradation of plasma facing material (PFM), however, as a layer joining PFM and cooling tube, the soft copper interlayer suffers concentrated thermal stress loads due to mismatched thermal expansion of PFM and cooling tube. Its thermal fatigue behavior under such coupled loads is also of great significance to structural safety of divertor component. With such a motivation, the reduction effects on fatigue life time of a typical interlayer of monoblock divertor under series of coupled HHF and ELMs shock loading conditions are investigated. It is found that: (1) The transient shock feature of ELMs loading is propagated into interlayer with less sharp pattern. The increase of damage induced by coupled ELMs loading is limited in single cycle, while the accumulated damage of multiple consecutive coupled loading cycles is increased nonlinearly. (2) Under the coupled HHF and ELMs loading, the fatigue life time of interlayer is generally decreasing. The magnitude of decrease is increasing nonlinearly with the magnitude of ELMs peak and averaged heat flux. (3) For three characteristic parameters of ELMs shock loading such as frequency, duration and peak heat flux, the peak heat flux and frequency are two parameters more sensitive to determine coupled reduction effects on fatigue lifetime of the interlayer, while for high frequency case, time averaged heat flux takes the lead.     

Analysis, verification, and benchmarking of the transient thermal hydraulic ITERTHA code for the design of ITER divertor

An analytical study for the International Thermonuclear Experimental Reactor Thermal Hydraulic Analysis code (ITERTHA) is carried out for a copper divertor with a 5 mm tungsten tile. The influence of the incident heat flux, swirl-tape insertion in cooling channels as well as the coolant flow velocity on the divertor thermal response is analyzed and discussed. The ITERTHA code results are verified by the commercial finite element code, COSMOS. The heat transfer coefficients at the nodes located on the cooling channel-wall are determined outside COSMOS code by the same methodology used in ITERTHA. A good agreement is achieved under different incident heat fluxes. The ITERTHA code is also benchmarked against the thermal-hydraulic calculation of the outer divertor of the Fusion Ignition Research Experiment, FIRE for an incident heat flux of 20 MW/m² and coolant flow velocity of 10 m/s in a cooling channel of 8 mm diameter with swirl-tape inserts of 2 ratio and 1.5 mm thickness. The results show excellent agreement for both steady and transient states and prove the successful implementation of both the hydraulic and heated diameters of the swirl-tape channels in the used heat transfer correlations.     

Simulation of Toroidal Rotation Effect on Heat Flux Transport in the Edge Plasma of Small Size Divertor Tokamak

The effect of toroidal rotation on heat flux transport in the edge plasma of small size divertor was simulated by B2SOLP0.5.2D transport code. The main results of simulation shows that, the following: (1) the radial heat flux is strongly influenced by toroidal rotation. (2) The amplification of conduction part of radial heat flux imposes nonresilient profile of ion temperature, under which the effect of toroidal rotation on ion temperature profile is strong. (3) The ion distribution and its gradients are lower for counter-injection neutral beam than for co-injection neutral beam. (4) Reversal of toroidal rotation during using neutral beam injection result in reverses of radial electric field and E × B drift velocity. (5) The toroidal rotation strong influence on the ion temperature scale length of the ion temperature gradient (ITG). (6) Switch on and off all drifts leads to higher change in the ion density distribution in edge plasma of small size divertor tokamak when the unbalance neutral beam injection are considered (7) the comparison between radial heat flux at different momentum input shows that, the radial ion heat flux with larger ion temperature scale length in the case of co-injection neutral beam is larger than the ion heat flux with smaller ion temperature scale length in the case of counter-injection neutral beam.     

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.     

Conceptual design and heat transfer performance of a flat-tile water-cooled divertor target

The divertor target components for the Chinese fusion engineering test reactor (CFETR) and the future experimental advanced superconducting tokamak (EAST) need to remove a heat flux of up to ～20 MW m-2.In view of such a high heat flux removal requirement,this study proposes a conceptual design for a flat-tile divertor target based on explosive welding and brazing technology.Rectangular water-cooled channels with a special thermal transfer structure (TTS)are designed in the heat sink to improve the flat-tile divertor target's heat transfer performance(HTP).The parametric design and optimization methods are applied to study the influence of the TTS variation parameters,including height (H),width (W*),thickness (T),and spacing (L),on the HTP.The research results show that the flat-tile divertor target's HTP is sensitive to the TTS parameter changes,and the sensitivity is T ＞ L ＞ W* ＞ H.The HTP first increases and then decreases with the increase of T,L,and W* and gradually increases with the increase of H.The optimal design parameters are as follows:H =5.5 mm,W* =25.8 mm,T =2.2 mm,and L =9.7 mm.The HTP of the optimized flat-tile divertor target at different flow speeds and tungsten tile thicknesses is studied using the numerical simulation method.A flat-tile divertor mock-up is developed according to the optimized parameters.In addition,high heat flux (HHF)tests are performed on an electron beam facility to further investigate the mock-up HTP.The numerical simulation calculation results show that the optimized flat-tile divertor target has great potential for handling the steady-state heat load of 20 MW m-2 under the tungsten tile thickness＜5 mm and the flow speed ≥7 m s-1.The heat transfer efficiency of the flat-tile divertor target with rectangular cooling channels improves by ～ 13％ and ～30％ compared to that of the flat-tile divertor target with circular cooling channels and the ITER-like monoblock,respectively.The HHF tests indicate that the flat-tile divertor mock-up can successfully withstand 1000 cycles of 20 MW m-2 of heat load without visible deformation,damage,and HTP degradation.The surface temperature of the flat-tile divertor mock-up at the 1000th cycle is only ～930 ℃.The fiat-tile divertor target's HTP is greatly improved by the parametric design and optimization method,and is better than the ITER-like monoblock and the fiat-tile mock-up for the WEST divertor.This conceptual design is currently being applied to the engineering design of the CFETR and EAST flat-tile divertors.     

FEM investigation and thermo-mechanic tests of the new solid tungsten divertor tile for ASDEX Upgrade

A new solid tungsten divertor for the fusion experiment ASDEX Upgrade is under construction at present. A new divertor tile design has been developed to improve the thermal performance of the current divertor made of tungsten coated fine grain graphite. Compared to thin tungsten coatings, divertor tiles made of massive tungsten allow to extend the operational range and to study the plasma material interaction of tungsten in more detail. The improved design for the solid tungsten divertor was tested on different full scale prototypes with a hydrogen ion beam. The influence of a possible material degradation due to thermal cracking or recrystallization can be studied. Furthermore, intensive Finite Element Method (FEM) numerical analysis with the respective test parameters has been performed. The elastic–plastic calculation was applied to analyze thermal stress and the observed elastic and plastic deformation during the heat loading. Additionally, the knowledge gained by the tests and especially by the numerical analysis has been used to optimize the shape of the divertor tiles and the accompanying divertor support structure.This paper discusses the main results of the high heat flux tests and their numerical simulations. In addition, results from some special structural mechanic analysis by means of FEM tools are presented. Finally, first results from the numerical lifecycle analysis of the current tungsten tiles will be reported.     

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.