Design of a 3D-printed liquid lithium divertor target plate and its interaction with high-density plasma

A liquid Li divertor is a promising alternative for future fusion devices.In this work a new divertor model is proposed,which is processed by 3D-printing technology to accurately control the size of the internal capillary structure.At a steady-state heat load of 10 MW m^(-2),the thermal stress of the tungsten target is within the bearing range of tungsten by finite-element simulation.In order to evaluate the wicking ability of the capillary structure,the wicking process at 600℃ was simulated by FLUENT.The result was identical to that of the corresponding experiments.Within 1 s,liquid lithium was wicked to the target surface by the capillary structure of the target and quickly spread on the target surface.During the wicking process,the average wicking mass rate of lithium should reach 0.062 g s^(-1),which could even supplement the evaporation requirement of liquid lithium under an environment>950℃.Irradiation experiments under different plasma discharge currents were carried out in a linear plasma device(SCU-PSI),and the evolution of the vapor cloud during plasma irradiation was analyzed.It was found that the target temperature tends to plateau despite the gradually increased input current,indicating that the vapor shielding effect is gradually enhanced.The irradiation experiment also confirmed that the 3D-printed tungsten structure has better heat consumption performance than a tungsten mesh structure or multichannel structure.These results reveal the application potential and feasibility of a 3D-printed porous capillary structure in plasma-facing components and provide a reference for further liquid-solid combined target designs.

Calculation and prediction of divertor detachment via impurity seeding by using one-dimensional model

Achieving the detachment of divertor can help to alleviate excessive heat load and sputtering problems on the target plates,thereby extending the lifetime of divertor components for fusion devices.In order to provide a fast but relatively reliable prediction of plasma parameters along the flux tube for future device design,a one-dimensional(1D)modeling code for the operating point of impurity seeded detached divertor is developed based on Python language,which is a fluid model based on previous work(Plasma Phys.Control.Fusion 58045013(2016)).The experimental observation of the onset of divertor detachment by neon(Ne)and argon(Ar)seeding in EAST is well reproduced by using the 1D modeling code.The comparison between the 1D modeling and two-dimensional(2D)simulation by the SOLPS-ITER code for CFETR detachment operation with Ne and Ar seeding also shows that they are in good agreement.We also predict the radiative power loss and corresponding impurity concentration requirement for achieving divertor detachment via different impurity seeding under high heating power conditions in EAST and CFETR phase II by using the 1D model.Based on the predictions,the optimized parameter space for divertor detachment operation on EAST and CFETR is also determined.Such a simple but reliable 1D model can provide a reasonable parameter input for a detailed and accurate analysis by 2D or three-dimensional(3D)modeling tools through rapid parameter scanning.

Simulation of tungsten impurity transport by DIVIMP under different divertor magnetic configurations on HL-3

Tungsten(W)accumulation in the core,depending on W generation and transport in the edge region,is a severe issue in fusion reactors.Compared to standard divertors(SDs),snowflake divertors(SFDs)can effectively suppress the heat flux,while the impact of magnetic configurations on W core accumulation remains unclear.In this study,the kinetic code DIVIMP combined with the SOLPS-ITER code is applied to investigate the effects of divertor magnetic configurations(SD versus SFD)on W accumulation during neon injection in HL-3.It is found that the W concentration in the core of the SFD is significantly higher than that of the SD with similar total W erosion flux.The reasons for this are:(1)W impurities in the core of the SFD mainly originate from the inner divertor,which has a short leg,and the source is close to the divertor entrance and upstream separatrix.Furthermore,the W ionization source(S_(W0))is much stronger,especially near the divertor entrance.(2)The region overlap of S_(W0)and F_(W,TOT)pointing upstream promote W accumulation in the core.Moreover,the influence of W source locations at the inner target on W transport in the SFD is investigated.Tungsten impurity in the core is mainly contributed by target erosion in the common flux region(CFR)away from the strike point.This is attributed to the fact that the W source at this location enhances the ionization source above the W ion stagnation point,which sequentially increases W penetration.Therefore,the suppression of far SOL inner target erosion can effectively prevent W impurities from accumulating in the core.

Closed corner divertor with B×▽B away from the divertor:a promising divertor scenario for tokamak power exhaust

A major challenge facing the steady-state operation of tokamak fusion reactors is to develop a viable divertor solution with order-of-magnitude increase in power handling capability as compared with present experience.A recently developed divertor concept for this end has been tested recently on EAST tokamak through combining the effects of a closed divertor corner and E×B drifts.The E×B drifts in the divertor move particles towards the outer divertor corner area in the scrape-off layer for B×▽B directed away from the divertor,which can significantly enhance the particle concentration there,facilitating divertor detachment.In recent EAST experiments,the effects have been demonstrated where the lowest electron temperature at the divertor plate is obtained with strike point located close to the corner in the horizontal target and with B×▽B away from the divertor.These experimental results are in reasonable agreement with SOLPS-ITER simulations including drift effects,suggesting that the new divertor concept potentially provides a promising divertor solution for long-pulse operations of future tokamak fusion reactors with much higher power fluxes.

Improvement of a temperature response function for divertor heat flux monitoring in fusion devices

Temperature response functions have been developed to investigate sensor design and divertor heat flux estimation in magnetically confined plasmas. The time-dependent heat flux can be derived by fitting the response function to experimental thermocouple(TC) data. Because the TC signals have a time delay to transit events such as discharge start or confinement transition, the time delay is taken into account in a temperature response function. Such a function accurately describes the signal from each TC channel with time delay in a sensor test using a neutral beam injection. Measurement for commercial TCs shows that the time delay is caused by the finite heat capacity of TC wire and contact heat resistance between TC and target surface.

核聚变装置偏滤器靶板材料选择与研究进展

可控热核聚变能是人类最理想的清洁能源之一,是解决人类能源和环境问题的根本途径。目前,可控热核聚变能的发展面临诸多挑战,偏滤器靶板作为磁约束核聚变装置中的重要部件,其设计和制造是维持等离子体放电、实现核聚变反应堆稳定运行亟需解决的关键问题之一。作为等离子体轰击最严重区域,偏滤器靶板经受着高能粒子流辐照和高热负荷冲击(10 MW·m^(-2)稳态负荷和20 MW·m^(-2)瞬态负荷),同时承担着磁约束聚变装置最主要的排热功能。因此,研发具有优异性能的靶板材料是推动核聚变能发展的关键一步。目前,金属铍、碳基材料以及钨基材料是主要的3种偏滤器靶板候选材料。基于国内外现有研究成果,论述了偏滤器靶板材料的选择与研究进展,对比分析了3种候选材料的优势以及存在的问题,以期为偏滤器靶板材料的选择、研发提供借鉴。

高导热W-金刚石复合材料钨靶模块热应力分析

核聚变堆中偏滤器在高热通量条件下服役将产生高温及高热应力导致钨靶模块开裂失效。探究了自主研发的新型W-金刚石复合材料不同热导率对偏滤器钨靶模块温度及热应力的预期影响。首先通过理论公式计算了W-金刚石复合材料的理论物性,并利用ANSYS软件模拟对比了以不同热导率的W-金刚石复合材料和纯W材料作为面向等离子体材料(PFM)的钨靶模块在20 MW/m^(2)热通量下的温度分布及热应力。结果表明,采用W-金刚石复合材料作为PFM能够极大程度地降低模块表面的最高温度,改善模块温度梯度,降低模块等效应变,但由于复合材料弹性模量较高,模块的等效应力并未得到缓解。该结果将对材料的设计及偏滤器结构优化提供重要的技术支持。

Soret和Dufour效应耦合对HL-2A偏滤器靶板氘浓度与温度双扩散的影响

在L/H模放电情况下,考虑Soret和Dufour效应耦合影响,用JD-CF程序研究了HL-2A偏滤器靶板中氘粒子浓度、温度、滞留量随时间变化影响。结果表明,在L/H模放电下,Soret和Dufour效应会对不同靶板材料(RAFM钢/钨)内不同位置处氘浓度分布、温度分布、氘滞留量产生一定影响,其最大变化率约为10%。在H模放电下Soret和Dufour效应对靶板的影响比在L模放电下的更明显,且对RAFM钢的影响比对钨的更明显。该研究结果对研究边缘等离子体输运、聚变堆燃料再循环、氚的滞留及投料量有一定参考价值。

偏滤器部件中W/Cu模块传热的数值模拟分析

利用理论公式和CFX仿真分别对偏滤器部件中平板型W/Cu模块进行稳态热分析,获得了不同热流密度下W/Cu模块的温度场和应力场的变化。理论计算得到的对流传热系数约为52304W·(m^(2)·K)^(‒1),而用CFX仿真分析得到的为35168~52186W·(m^(2)·K)^(‒1)。此外,两种稳态热分析方式得到的W/Cu模块的温度场数值相近;在20MW·m^(‒2)热流密度下,W/Cu模块的温度场、应力场分析结果均超出了对应材料的许用范围,严重影响了偏滤器部件的稳定运行。

Characteristics of the SOL ion-to-electron temperature ratio on the J-TEXT tokamak with different plasma configurations

Accurate measurement of the average plasma parameters in the edge region,including the temperature and density of electrons and ions,is critical for understanding the characteristics of the scrape-off layer(SOL) and divertor plasma transport in magnetically confined fusion research.On the J-TEXT tokamak,a multi-channel retarding field analyzer(RFA) probe has been developed to study average plasma parameters in the edge region under various poloidal divertor and island divertor configurations.The edge radial profile of the ion-to-electron temperature ratio,τ_(i/e),has been determined,which gradually decreases as the SOL ion self-collisionality,v_(SOL)*,increases.This is broadly consistent with what has been observed previously from various tokamak experiments.However,the comparison of experimental results under different configurations shows that in the poloidal divertor configuration,even under the same v_(SOL)*,τ_(i/e) in the SOL region becomes smaller as the distance from the X-point to the target plate increases.In the island divertor configuration,τ_(i/e) near the O-point is higher than that near the X-point at the same v_(SOL)*,and both are higher than those in the limiter configuration.These results suggest that the magnetic configuration plays a critical role in the energy distributions between electrons and ions at the plasma boundary.

HL-3先进雪花偏滤器位形的设计与敏感性分析

使用EFIT程序设计了HL-3装置雪花偏滤器位形,研究了雪花位形对HL-3装置极向场(PF)线圈电流变化的敏感性。研究发现:当PF线圈电流变化达1%时,标准雪花偏滤器转变为雪花加位形或是雪花减位形;当PF线圈电流变化超过8.8%时,雪花减位形在大偏离度下能够获得平衡位形,而雪花加位形难以获得收敛结果。

HL-2A托卡马克偏滤器脱靶时边缘极向旋转和湍流动量输运

偏滤器脱靶为降低托卡马克靶板热负荷提供了一种有效的解决方案,但脱靶可能引起边界等离子体状态发生变化,影响整体约束性能.本文报道了在中国环流器二号A托卡马克上开展的L模放电偏滤器脱靶时边界等离子体极向旋转和湍流动量输运的实验研究.采用在偏滤器室注入混合气体(60%氮气+40%氘气)的方式实现了偏滤器脱靶.研究发现,在未脱靶-预脱靶-脱靶过程中,实验测得的近刮削层区域E×B极向流速与湍流动量对极向旋转的驱动作用(雷诺应力)的演化一致;相较于未脱靶状态,脱靶时等离子体边缘极向速度剪切明显降低,导致湍流水平增强.在湍流输运和辐射都增强的共同作用下,等离子体整体约束性能下降.研究表明,边缘湍流输运和等离子体旋转动力学在偏滤器脱靶影响整体约束的芯-边耦合机制中发挥作用.

CFETR偏滤器低温泵抽气系统的设计和有效抽速的计算

中国聚变工程实验堆(CFETR)提出了混合偏滤器-包层集成概念,偏滤器下方被增殖包层占据。偏滤器抽气通道设计在增殖包层内,满足有效抽气流量为250Pa·m^(3)·s^(-1)的工程设计要求。基于真空流导计算理论设计了偏滤器低温泵抽气系统,计算比较了不同截面形状下偏滤器低温泵抽气系统的流导和单台低温泵在偏滤器抽气通道入口的有效抽速。在增殖包层内设计正方形截面形状的抽气通道更为合理,可同时满足粒子排除要求和增殖包层强度要求。当正方形截面尺寸为350mm×350mm,偏滤器低温泵抽气系统流导为12.08m^(3)·s^(-1),单台低温泵在偏滤器抽气通道入口的有效抽速为9.905m^(3)·s^(-1)时,26台低温泵可满足粒子排除要求,假设一个下窗口设置2台低温泵,低温泵共占用13个下窗口。基于理论公式设计了偏滤器低温泵抽气系统和评估了偏滤器的粒子排除能力,为后续CFETR的偏滤器抽气通道优化和有效抽速评估提供了理论依据和参考。

EAST托卡马克钨杂质上下不对称性分布的实验研究

基于EAST托卡马克装置高性能极紫外空间分辨杂质光谱仪,本文首次开展了磁约束聚变装置高Z杂质上下不对称分布的实验研究.实验结果显示,在同向中性束注入期间,等离子体芯部环向旋转速度V_(t0)大,钨杂质上下不对称性较强,且辐射(密度)较强的一侧背离离子B×▽B漂移方向;当从上外偏滤器充气口注入氘化甲烷气体后,V_(t0)迅速下降,原有上下不对称性发生反转.因此我们针对近似条件下W^(32+)杂质离子特征线辐射不对称因子I_(u)/I_(d)与V_(t0)依赖关系开展了进一步的统计分析.结果表明,I_(u)/I_(d)正相关于V_(t0),当V_(t0)<20 km/s以下时,不对称性发生反转.上述现象从实验角度验证了漂移动理学的理论预测,说明环向旋转带来的离心力影响了杂质离子平行于磁场方向的动量守恒,作为直接诱因,造成了高Z杂质密度的上下不对称分布,进而影响辐射的分布.本文对钨杂质上下不对称性分布的实验观测为进一步开展高Z杂质极向输运的机理研究打下了坚实的基础,并为今后聚变堆高Z杂质控制提供重要的参考.

Review of Divertor Studies in LHD

In the Large Helical Device （LHD）, two different divertor configurations, i.e. helical divertor （HD） and local island divertor （LID）, are utilized to control the edge plasma. The HD with two X-points is an intrinsic divertor for heliotron devices, accompanied with a relatively thick ergodic layer outside the confinement region. Edge and divertor plasma behavior from low density to high density regimes is presented, referring to the divertor detachment. The effect of the ergodic layer on the edge transport is also discussed. On the other hand, the LID is an advanced divertor concept which realizes a high pumping efficiency by the combination of an externally induced magnetic island and a closed pumping system. Experimental results to confirm the fundamental divertor performance of the LID are presented.

Calculation of the heat flux in the lower divertor target plate using an infrared camera diagnostic system on the experimental advanced superconducting tokamak

During the discharging of Tokamak devices, interactions between the core plasma and plasma-facing components (PFCs) may cause exorbitant heat deposition in the latter. This poses a grave threat to the lifetimes of PFCs materials. An infrared (IR) diagnostic system consisting of an IR camera and an endoscope was installed on an Experimental Advanced Superconducting Tokamak (EAST) to monitor the surface temperature of the lower divertor target plate (LDTP) and to calculate the corresponding heat flux based on its surface temperature and physical structure, via the finite element method. First, the temperature obtained by the IR camera was calibrated against the temperature measured by the built-in thermocouple of EAST under baking conditions to determine the true temperature of the LDTP. Next, based on the finite element method, a target plate model was built and a discretization of the modeling domain was carried out. Then, a heat conduction equation and boundary conditions were determined. Finally, the heat flux was calculated. The new numerical tool provided results similar to those for DFLUX;this is important for future work on related physical processes and heat flux control.

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.

The first results of divertor discharge and supersonic molecular beam injection on the HL-2A tokamak

HL-2A tokamak is the first tokamak with divertors in China. The plasma boundary and the position of the striking point on the target plates of the HL-2A closed diwrtor were simulated by the current filament code and they were in agreement with the diagnostic results in the divertor. Supersonic molecular beam injection （SMBI） system was first installed and tested on the HL-2A tokamak in 2004. In the present experiment low pressure SMBI fuelling on the HL-2A closed divertor was carried out. The experimental results indicate that the divertor was operated in the ＇linear regime＇ and during the period of SMB pulse injection into the HL-2A plasma the power density eonvected at the target plate surfaces was 0.4 times of that before or after the beam injection. It is a useful fuelling method for decreasing the heat load on the neutralizer plates of the divertor.

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.

Study of Striated Heat Flux on EAST Divertor Plates Induced by LHW Using Infrared Camera

An upgraded infrared （IR） imaging system which provides a wide field of view （FOV） has been installed on the Experimental Advanced Superconducting Tokamak （EAST） to monitor the surface temperatures on plasma facing components. Modified magnetic topology induced by lower hybrid wave （LHW） can lead to the formation of striated heat flux （SHF} on divertor plates which can be clearly observed by IR camera. In this paper, LHW power modulation is applied to analyze the appearance of SHF. It is also demonstrated that deuterium （D） pellet injection and supersonic molecular beam injection （SMBI） can to some extent reduce the heat flux on the outer strike point （OSP）, but enhance the SHF on lower outer plates （LOP） of divertor. This may provide an optional approach to actively control the distribution of heat flux on diveror plates, which can protect materials from long duration high-heat flux.