Realization of minute-long steady-state H-mode discharges on EAST

In the 2016 EAST experimental campaign,a steady-state long-pulse H-mode discharge with an ITER-like tungsten divertor lasting longer than one minute has been obtained using only RF heating and current drive,through an integrated control of the wall conditioning,plasma configuration,divertor heat flux,particle exhaust,impurity management,and effective coupling of multiple RF heating and current drive sources at high injected power.The plasma current (Ip ～ 0.45 MA) was fully-noninductively driven (Vloop ＜ 0.0 V) by a combination of ～2.5 MW LHW,～0.4 MW ECH and ～0.8 MW ICRF.This result demonstrates the progress of physics and technology studies on EAST,and will benefit the physics basis for steady state operation of ITER and CFETR.     

Realization of divertor configuration discharge in J-TEXT tokamak

To extend the operation region of the Joint-Texas Experimental tokamak (J-TEXT) to the divertor configuration and even the H-mode, the divertor configuration discharge has been realized for the first time in the J-TEXT tokamak. Along with the establishment of a power supply for the divertor configuration, the construction of relevant diagnostics, and the installation of the divertor target on the high-field side, divertor discharge has been tested. Through the equilibrium calculation and position stability analysis, the control strategy has evolved to be more stable. High-density experiments and auxiliary heating experiments have been carried out on the divertor configuration. The special midplane single-null (MSN) divertor configuration is shown to be more stable than the limiter configuration in the density limit condition and can reach a higher density in the experiment. In the ECRH experiment, the power injection enhances the electron temperature and density, while more heat outflux is loaded on the divertor target tiles and causes more intensive recycling and impurity release. The future plan for the divertor configuration operation in the J-TEXT tokamak is also included.     

Lithium coatings on NSTX plasma facing components and its effects on boundary control, core plasma performance, and operation

NSTX high power divertor plasma experiments have used in succession lithium pellet injection (LPI), evaporated lithium, and injected lithium powder to apply lithium coatings to graphite plasma facing components. In 2005, following the wall conditioning and LPI, discharges exhibited edge density reduction and performance improvements. Since 2006, first one, and now two lithium evaporators have been used routinely to evaporate lithium onto the lower divertor region at total rates of 10-70 mg/min for periods 5-10 min between discharges. Prior to each discharge, the evaporators are withdrawn behind shutters. Significant improvements in the performance of NBI heated divertor discharges resulting from these lithium depositions were observed. These evaporators are now used for more than 80% of NSTX discharges. Initial work with injecting fine lithium powder into the edge of NBI heated deuterium discharges yielded comparable changes in performance. Several operational issues encountered with lithium wall conditions, and the special procedures needed for vessel entry are discussed. The next step in this work is installation of a liquid lithium divertor surface on the outer part of the lower divertor.     

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。     

Design and performance study of the helium-cooled T-tube divertor concept

The ARIES-CS study has been launched with the goal of developing through physics and engineering optimization an attractive power plant concept based on a compact stellarator configuration. The study included an effort to characterize the divertor location and corresponding heat load distribution, and to develop a He-cooled divertor concept that could accommodate a heat flux of at least 10 MW/m², and that would integrate well with the other power core components. This paper describes the design study of this divertor concept, which, although developed for a compact stellarator, is well suited for a tokamak configuration also.     

Integrated Operating Scenario to Achieve 100-Second,High Electron Temperature Discharge on EAST

Stationary long pulse plasma of high electron temperature was produced on EAST for the first time through an integrated control of plasma shape,divertor heat flux,particle exhaust,wall conditioning,impurity management,and the coupling of multiple heating and current drive power.A discharge with a lower single null divertor configuration was maintained for 103 s at a plasma current of 0.4 MA,q_(95)≈7.0,a peak electron temperature of 4.5 keV,and a central density n_e(0)~2.5×10~(19) m~(-3).The plasma current was nearly non-inductive(V_(loop) 0.05 V,poloidal beta ~0.9) driven by a combination of 0.6 MW lower hybrid wave at 2.45 GHz,1.4 MW lower hybrid wave at 4.6 GHz,0.5 MW electron cyclotron heating at 140 GHz,and 0.4 MW modulated neutral deuterium beam injected at 60 kV.This progress demonstrated strong synergy of electron cyclotron and lower hybrid electron heating,current drive,and energy confinement of stationary plasma on EAST.It further introduced an example of integrated "hybrid" operating scenario of interest to ITER and CFETR.     

Research plan for divertor simulation making use of a large tandem mirror device

We are planning to start a study of divertor simulation under the closely resemble to actual fusion plasma environment making use of the advantage of open magnetic field configuration and to contribute the solution for realizing the divertor in ITER as a future research plan of Plasma Research Center of the University of Tsukuba. In the research plan, the concepts of two divertor devices are introduced. One has an axi-symmetric divertor configuration with the separatrix which is similar to toroidal divertor of torus systems and the other is a high heat flux divertor simulator by using an end-mirror exit of the existing tandem mirror device. Development of magnetic field configuration for ensuring the MHD stability is under way and a designed example is investigated under the optimal condition for plasma production. Consideration of plasma heating scheme using Fokker-Planck simulation code was successfully performed at both axi-symmetric divertor and end-mirror regions. Preparative experiments using calorimeter, Mach probe and high-speed camera have been started at the end-mirror region and the heat flux density of the level in 1-10 MW m⁻² was achieved in standard hot-ion mode plasma-confining experiments, which gives a clear prospect of generating the required heat flux density for divertor studies.     

Improvement of divertor triple probe system and its measurements under full graphite wall on EAST

Full graphite wall of experimental advanced superconducting tokamak (EAST) has been developed in the spring of 2008. A new divertor triple probe diagnostics system (DTPDs) is built for EAST during this upgrade. The tip shape and connected structure of the probe are optimized for variational magnetic field directions and DTPDs maintenance. The experiment has been carried out with a full graphite wall for EAST, and near double-null diverted plasma is achieved successfully. The evolutions of electron temperature, density, particle flux and power densities along the divertor targets have been obtained with DTPDs.     

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.     

The design, analysis and alignment of EAST divertor

The divertor of EAST is composed of inner-target, outer-target and dome. Every kind of target has 16 separate modules along toroidal direction. The structure of divertor includes support, heat sink and the first wall graphite. The shape and dimension of divertor are optimized for flexible operation of different elongation and triangularity of plasma. The eddy current and halo current interact with high toroidal magnetic field will produce large forces on divertor, so the structure of divertor must bear these electromagnetic forces. The first wall of divertor must have high assembly accuracy of 0.5 mm. The traditional method using optical instruments to survey directly is impractical because of the especial location of divertor. The benchmark templet integrates with optical instruments to align the divertor is proved simple and effective.     

Operational impact on the JET ITER-like wall in-vessel components

The JET ITER-like Wall (ILW) provides the same plasma facing component configuration as ITER during its active phase: beryllium in the main chamber and tungsten in the divertor. Moving from a carbon-based wall to an all metals wall requires some operational adjustment. The reduction in radiation at the plasma edge and in the divertor can lead to high power loads on the plasma facing components both in steady state and in transients and requires the development of radiative scenarios and the use of massive gas injection to mitigate disruptions. These tools are even more important now because an all metal wall is much less forgiving to thermal overloading the carbon based wall used to be. Here the impact of the first 11 months of operation on the ILW plasma facing components is discussed.     

Analysis of Interaction of Disruption Induced Runaway Electrons with Plasma Facing Components in Next Generation Tokamaks

Runaway electrons which are accelerated during plasma disruptions may cause damage to the plasma facing components when their energy is deposited locally. In order to assess the possible damage of plasma facing components and the associated damage thresholds in a next generation tokamak, analyses have been carried out. The energy deposition by 100 and 300 MeV electrons in component materials has been calculated using a Monte Carlo code. The effect of parametric changes of carbon armor thickness, electron energy and angle of incidence has been evaluated. Subsequently the thermal response of divertor structures with carbon armor and with bare tungsten, and of a first wall structure has been analyzed and thresholds for thermally induced component damage were derived. The damage threshold under 100 MeV electron impact on a divertor structure with 10 mm carbon coverage and dispersion strengthened copper cooling tubes is about 60 MJ/m² of incident energy, that for a divertor structure with molybdenum coolant tubes is about 115 MJ/m², whereas the damage threshold for melting of the bare tungsten divertor is only about 30 MJ/m². Damage of the first wall structure would occur above 180 MJ/m². For 300 MeV electron incidence the damage thresholds are 13 to 47% lower than the values for 100 MeV.     

Design of snowflake-diverted equilibria of CFETR

The Chinese Fusion Engineering Test Reactor (CFETR) represents the next generation of full superconducting fusion reactors in China.Recently,CFETR was redesigned with a larger size and will be operated in two phases.To reduce the heat flux on the target plate,a snowflake (SF) divertor configuration is proposed.In this paper we show that by adding two dedicated poloidal field (PF) coils,the SF configuration can be achieved in both phases.The equilibria were calculated by TEQ code for a range of self-inductances li3.The coil currents were calculated at some fiducial points in the flattop phase.The results indicate that the PF coil system has the ability to maintain a long flattop phase in 7.5 and 10 MA inductive scenarios for the single null divertor (SND) and SF divertor configurations.The properties of the SF configuration were also analyzed.The connection length and flux expansion of the SF divertor were both increased significantly over the SND.     

Double-null divertor configuration discharge and disruptive heat flux simulation using TSC on EAST

The tokamak simulation code (TSC) is employed to simulate the complete evolution of a disruptive discharge in the experimental advanced superconducting tokamak.The multiplication factor of the anomalous transport coefficient was adjusted to model the major disruptive discharge with double-null divertor configuration based on shot 61 916.The real-time feed-back control system for the plasma displacement was employed.Modeling results of the evolution of the poloidal field coil currents,the plasma current,the major radius,the plasma configuration all show agreement with experimental measurements.Results from the simulation show that during disruption,heat flux about 8 MW m-2 flows to the upper divertor target plate and about 6 MW m-2 flows to the lower divertor target plate.Computations predict that different amounts of heat fluxes on the divertor target plate could result by adjusting the multiplication factor of the anomalous transport coefficient.This shows that TSC has high flexibility and predictability.     

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.     

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.     

Vacuum System for HL-2A Tokamak

The vacuum system for HL-2A was built in 2003. The test results indicated that this system is feasible. It consists of three main parts: a pumping system, a pumping divertor and a glow discharge cleaning (GDC) system. For the pumping system, there are three main functions: (1) evacuating the vacuum vessel thus to produce an ultra high vacuum, (2) removal of impurities released during baking and (3) pumping during GDC. The pumping divertor controls the particles at the plasma edge and the GDC system provides a clean wall conditioning. During the first campaign of physical trial experiment on HL-2A, the ultimate pressure reached 4.6 × 10-6 Pa, and the total leakage and outgassing rate in 12 hours was 1.8 × 10-5 Pa·m3/s, which is close to that of ASDEX.     

Feasibility study of an actively cooled tungsten divertor in Tore Supra for ITER technology testing

In order to reduce the risks for ITER Plasma Facing Components (PFCs), it is proposed to equip Tore Supra with a full tungsten divertor, benefitting from the unique long pulse capabilities, the high installed RF power and the long experience with actively cooled high heat flux components of the Tore Supra platform. The transformation from the current circular limiter geometry to the required X-point configuration will be achieved by installing a set of copper poloidal coils inside the vacuum vessel. The new configuration will allow for H-mode access, providing relevant plasma conditions for PFC technology validation. Furthermore, attractive steady-state regimes are expected to be achievable. The lower divertor target design will be closely based on that currently envisaged for ITER (W monoblocks), while the upper divertor region will be used to qualify the main first wall heat sink technology adopted for the ITER blanket modules (CuCrZr copper/stainless steel) with a tungsten coating (in place of the Be tiles which ITER will use). Extended plasma exposure will provide access to ITER critical issues such as PFC lifetime (melting, cracking, etc.), tokamak operation on damaged metallic surfaces, real time heat flux control through PFC monitoring, fuel retention and dust production.     

Optimizing the overall configuration of a He-cooled W-alloy divertor for a power plant

A number of different He-cooled divertor configurations have been proposed for magnetic fusion energy (MFE) power plant application. They range in scale from a plate configuration with characteristic dimension of the order of 1 m, to the ARIES-CS T-tube configuration with characteristic dimension of the order of 10 cm, to the EU FZK finger concept with characteristic dimension of the order of 1.5 cm. All these designs utilize tungsten or tungsten alloy as structural material. This paper considers the characteristics of the different divertor configurations and proposes the possibility of optimizing the design by combining different configurations in an integrated design based on the anticipated divertor heat flux profile.     

Calculation of the Heat Flux on the First Wall During Disruption on Tokamak

Disruptions are the most dangerous instabilities in tokamak plasma. During plasma disruption, the large amounts of energy will be deposited on Plasma Facing Components (PFCs) which is a damaging threat for the divertor target and the first wall materials. Therefore, studying the characteristic of heat deposition on the first wall is very significant. The Infrared (IR) camera is an effective tool to measure the surface temperature profile on the first wall on the Experimental Advanced Superconducting Tokamak (EAST). With a finite difference method, the heat flux arrived to the divertor can be calculated from the surface temperature. However, the surface layer on the divertor has a great influence on the calculation of the heat flux on the divertor. The numerical method for solving heat conduction for semi-infinite model is given in this paper. And the thermal resistance of surface layers is considered in this numerical method. In addition, the distribution of heat flux on the divertor during disruption is also shown.