Development of vacuum ultraviolet spectroscopy for measuring edge impurity emission in the EAST tokamak

The dominant wavelength range of edge impurity emissions moves from the visible range to the vacuum ultraviolet(VUV) range, as heating power increasing in the Experimental Advanced Superconducting Tokamak(EAST). The measurement provided by the existing visible spectroscopies in EAST is not sufficient for impurity transport studies for high-parameters plasmas. Therefore, in this study, a VUV spectroscopy is newly developed to measure edge impurity emissions in EAST. One Seya-Namioka VUV spectrometer(McPherson 234/302) is used in the system, equipped with a concave-corrected holographic grating with groove density of 600 grooves mm~(–1). Impurity line emissions can be observed in the wavelength range ofλ=50–700 nm, covering VUV, near ultraviolet and visible ranges. The observed vertical range is Z=-350–350 mm. The minimum sampling time can be set to 5 ms under full vertical binning(FVB) mode. VUV spectroscopy has been used to measure the edge impurity emission for the 2019 EAST experimental campaign. Impurity spectra are identified for several impurity species, i.e., lithium(Li), carbon(C), oxygen(O), and iron(Fe). Several candidates for tungsten(W) lines are also measured but their clear identification is very difficult due to a strong overlap with Fe lines. Time evolutions of impurity carbon emissions of CII at 134.5 nm and CIII at97.7 nm are analyzed to prove the system capability of time-resolved measurement. The measurements of the VUV spectroscopy are very helpful for edge impurity transport study in the high-parameters plasma in EAST.     

DIVIMP-B2-EIRENE modelling of 13C migration and deposition in ASDEX Upgrade L-mode plasmas

Carbon transport and migration were studied experimentally and numerically in a high-density, low-confinement mode plasma in the ASDEX Upgrade tokamak. On the last day of plasma operation of the 2004–2005 experimental campaign, ¹³CH₄ was injected into the vacuum vessel from the low field side midplane. A poloidal set of tiles was subsequently removed and analysed for ¹³C deposition. In this work the measured deposition profile is interpreted using the impurity transport code DIVIMP. The simulated poloidal distribution of ¹³C deviates significantly from the measured profile. The simulations indicate that ¹³C is promptly deposited at the wall in the vicinity of the injection port, and is transported to the low field side divertor plate predominately via the scrape-off layer. The B2-EIRENE plasma solution produce stagnant plasma flow in the main scrape-off layer, in contrast to measurements in ASDEX Upgrade and other tokamaks. This is the likely cause of the discrepancy between the measured and the calculated poloidal distribution of the ¹³C deposition. Key model parameters of DIVIMP were varied to determine their effect on the simulated deposition profile.     

托卡马克等离子体杂质密度的光谱法测量研究

本文介绍了基于托卡马克等离子体被动光谱诊断获得杂质密度的方法。通过被动光谱诊断测量获得杂质线辐射的空间多道弦积分强度分布,利用强度标定系数转换为绝对光亮度分布;通过测量弦与等离子体位形,将弦积分的强度分布反演变换为径向体发射率。根据线辐射强度激发截面求出对应电离态的离子密度,最后采用杂质输运程序模拟计算得出总密度分布。以东方超环(Experimental Advanced Superconducting Tokamak,EAST)托卡马克装置上软X射线-极紫外光谱(Soft X-ray and Extreme Ultraviolet Spectrometers,XEUV)诊断测量到的Mo XXIX-Mo XXXII为例,描叙了获得Mo杂质密度分布的过程,获得的总误差小于10%。     

Predictive Modeling for Performance Assessment of ITER-Like Divertor in China Fusion Engineering Testing Reactor

To facilitate the design of the China Fusion Engineering Testing Reactor (CFETR), predictive modeling for the assessment and optimization of the divertor performances is an indispensable approach. This paper presents the modeling of the edge plasma behaviors as well as the W erosion and transport properties in CFETR with ITER-like divertor by using the B2-Eirene/SOLPS 5.0 code package together with the Monte Carlo impurity transport code DIVIMP. As expected, SOLPS modeling of divertor-SOL plasmas finds that the peak heat flux onto the divertor targets greatly exceeds 10 MW/m², an engineering limit posed to the steady-state and/or long-pulse operation of the next-step fusion devices, for a wide range of plasma conditions, and thus modeling of Ar puffing by scanning the puffing rate for radiative divertor is performed. As the increase of the Ar puffing rate, the peak target heat fluxes and plasma temperature decreases exponentially,reflecting that Ar puffing is highly effective at power exhausting. Based on the ion fluxes from SOLPS, the W erosion is calculated by taking into consideration the bombardment of both D and Ar ions, and then the W plasma concentrations are calculated based on the W erosion fluxes using DIVIMP. The calculations show that if the Ar puffing only being used to reduce the divertor heat load, the W plasma contamination in the core plasma exceeds the tolerable value (<10^?5), which demonstrates that some further upgrading of the divertor geometry is still needed.     

Calculation of Neutral Beam Injected Torque and Its Effective Tangency Major Radius for EAST

Toroidal rotation has been recognized to have significant effects on the transport and magnetohydrodynarnic(MHD) stability of tokamak plasmas.Neutral beam injection(NBI) is the most effective rotation generation method on current,tokamak devices.To estimate the effective injected torque of the first neutral beam injection system on EAST,a simplified analytic method was derived.Calculated beam torque values were validated by those obtained from the NUBEAM code simulation.According to the results,for the collisional torque,the effective tangential radius for torque deposition is close to the beam tangency major radius.However,due to the dielectric property of tokamak plasma,the equivalent tangency major radius of the J×B torque is equal to the average major radius of the magnetic flux surface.The results will be useful for the research of toroidal momentum confinement and the experimental analysis of momentum transport related with NBI on EAST.     

Shutdown dose rate calculation code system and its application to EAST tokamak

For safe operation and maintenance of nuclear devices, it is very important to predict the dose rate distribution after shutdown. Based on the rigorous two-step (R2S) method, a new shutdown dose rate calculation code system was developed for nuclear devices with large dimension and complex geometries. By coupling the Monte Carlo particle transport calculation code MCNP with the activation simulation code FISPACT, the dose rate calculation codes integrate the functions of neutron transport calculation, activation calculation and decay gamma transport calculation. This code system was applied to the shutdown dose analysis for experimental advanced superconducting tokamak (EAST). Three-dimensional dose rate distributions of the EAST for different cooling times and spatial locations were obtained. The results can be used to assist radiation protection in EAST.     

Assessment of Radiation Damage to the Structural Material of EAST Tokamak

Radiation damage to structural material of fusion facilities is of high concern for safety. The superconducting tokamak EAST will conduct D-D plasma experiments with the neutron production of 1015 neutrons per second. To evaluate the material radiation damage a programme system has been devised with the Monte Carlo transport code MCNP-4C, the inventory code FISPACT99, a specific interface, and the fusion evaluated nuclear data library FENDL-2. The key nuclear responses, i.e. fast neutron flux, displacement per atom, and the helium and hydrogen production, are calculated for the structural material SS-316L of the first wall, and the vacuum vessel, using this programme. The results demonstrate that the radiation damage to the structural material is so little that it will not lead to any significant change of material properties according to the reference design. This indicates that there is a large potential space for EAST to test advanced operation regime from the viewpoint of structural material safety.     

Observation of Impurity Accumulation After Hydrogen Multi-Pellet Injection in Large Helical Device

Impurity accumulation is studied for neutral beam-heated discharges after hydrogen multi-pellet injection in Large Helical Device (LHD). Iron density profiles are derived from radial profiles of EUV line emissions of FeXV-XXIV with the help of the collisional-radiative model. A peaked density profile of Fe²³⁺ is simulated by using one-dimensional impurity transport code. The result indicates a large inward velocity of -6 m/s at the impurity accumulation phase. However, the discharge is not entirely affected by the impurity accumulation, since the concentration of iron impurity, estimated to be 3.3x10^-5 to the electron density, is considerably small. On the other hand, a flat profile is observed for the carbon density of C⁶⁺, which is derived from the Z_eff profile, indicating a small inward velocity of -1 m/s. These results suggest atomic number dependence in the impurity accumulation of LHD, which is similar to the tokamak result.     

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.     

Investigation of lithium as plasma facing materials on HT-7

First experiment of liquid lithium limiter was successfully carried out on HT-7 tokamak and a few positive results were obtained. The results showed that by using lithium limiter, specially liquid lithium limiter, Hα intensity reduced 20-30%, the emission of CIII and OV decreased about 10-20%, loop voltage had a slight decline, the core electron temperature slightly increased, the particle confinement time increased by a factor of 2, and the energy confinement time increased 20%. After lithium coating, the hydrogen recycling decreased, and core electron temperature increased significantly by a factor of 2. At the same time, after lithium coating, electron density of edge plasmas obviously decreased while electron temperature slightly increased. These encouraging results are very useful for further research of long tray lithium limiter on HT-7 and liquid divertor on EAST.     

Line identification of extreme ultraviolet(EUV) spectra from low-Z impurity ions in EAST tokamak plasmas

Extreme ultraviolet(EUV) spectra emitted from low-Z impurity ions in the wavelength range of10–500 ? were observed in Experimental Advanced Superconducting Tokamak(EAST)discharges. Several spectral lines from K-and L-shell partially ionized ions were successfully observed with sufficient spectral intensities and resolutions for helium, lithium, boron, carbon,oxygen, neon, silicon and argon using two fast-time-response EUV spectrometers of which the spectral intensities are absolutely calibrated based on the intensity comparison method between visible and EUV bremsstrahlung continua. The wavelength is carefully calibrated using wellknown spectra. The lithium, boron and silicon are individually introduced for the wall coating of the EAST vacuum vessel to suppress mainly the hydrogen and oxygen influxes from the vacuum wall, while the carbon and oxygen intrinsically exist in the plasma. The helium is frequently used as the working gas as well as the deuterium. The neon and argon are also often used for the radiation cooling of edge plasma to reduce the heat flux onto the divertor plate. The measured spectra were analyzed mainly based on the database of National Institute of Standards and Technology. As a result, spectral lines of He Ⅱ, Li Ⅱ–Ⅲ, B Ⅳ–Ⅴ, C Ⅲ–Ⅵ, O Ⅲ–Ⅷ, Ne Ⅱ–Ⅹ,Si Ⅴ–Ⅻ, and Ar Ⅹ–XVI are identified in EAST plasmas of which the central electron temperature and chord-averaged electron density range in T_(e0)=0.6–2.8 keV and n_e=(0.5–6.0)×10~(19) m~(-3), respectively. The wavelengths and transitions of EUV lines identified here are summarized and listed in a table for each impurity species as the database for EUV spectroscopy using fusion plasmas.     

Kinetic micro-instabilities in the presence of impurities in toroidal magnetized plasmas

The theoretical and numerical studies on kinetic micro-instabilities,including ion temperature gradient(ITG) driven modes,trapped electron modes(TEMs) in the presence of impurity ions as well as impurity modes(IMs),induced by impurity density gradient alone,in toroidal magnetized plasmas,such as tokamak and reversed-field pinch(RFP) are reviewed briefly.The basic theory for IMs,the electrostatic instabilities in tokamak and RFP plasmas are discussed.The observations of hybrid and coexistence of the instabilities are categorized systematically.The effects of impurity ions on electromagnetic instabilities such as ITG modes,the kinetic ballooning modes(KBMs) and kinetic shear Alfvén modes induced by impurity ions in tokamak plasmas of finite β(=plasma pressure/magnetic pressure) are analyzed.The interesting topics for future investigation are suggested.     

Reconstructions of velocity distributions from fast-ion D-alpha(FIDA) measurements on EAST

2D fast-ion velocity-space distributions have been reconstructed from two-view fast-ion D-alpha(FIDA) measurements on experimental advanced superconducting tokamak(EAST). To make up for the sparse data and incomplete velocity-space coverage with the dual-view, we use nonnegativity and null-measurements as prior information to reconstruct the velocity distribution in experiments with co-and counter-current neutral beam injection. An improved reconstructed fast-ion distribution is achieved by combining the existing O-and B-port FIDA measurements with the proposed A-port FIDA view. To further improve the reliability of FIDA-based reconstructions on EAST, based on real multi-view FIDA measurements on EAST in the near future, various bases will be studied further.     

Study on Radial Position of Impurity Ions in Core and Edge Plasma of LHD Using Space-Resolved EUV Spectrometer

Radial profiles of impurity ions of carbon, neon and iron were measured for high-temperature plasmas in large helical device (LHD) using a space-resolved extreme ultraviolet (EUV) spectrometer in the wavelength range of 60 to 400?. The radial positions of the impurity ions obtained are compared with the local ionization energies, Ei of these impurity ions and the electron temperatures TeZ there. The impurity ions with 0.3?Ei?1.0 keV are always located in outer region of plasma, i.e., 0.7?ρ?1.0, and those with Ei?0.3keV are located in the ergodic layer, i.e., 1.0?ρ?1.1, with a sharp peak edge., where ρ is the normalized radial position. It is newly found that TeZ is approximately equal to Ei for the impurity ions with Ei?0.3keV, whereas roughly half the value of Ei for the impurity ions with 0.3?Ei?1.0keV. It is known that TeZ is considerably lower than Ei in the plasma edge and approaches to Ei in the plasma core. Therefore, this result seems to originate from the difference in the transverse transport between the plasma edge at ρ?1.0 and the ergodic layer at ρ?1.0. The transverse transport is studied with an impurity transport simulation code. The result revealed that the difference appearing in the impurity radial positions can be qualitatively explained by the different values of diffusion coefficient, e.g., D=0.2 and 1.0m2/s, which can be taken as a typical index of the transverse transport.     

Study of the tungsten sputtering source suppression by wall conditionings in the EAST tokamak

The steady fusion plasma operation is constrained by tungsten(W) material sputtering issue in the EAST tokamak. In this work, the suppression of W sputtering source has been studied by advanced wall conditionings. It is also concluded that the W sputtering yield becomes more with increasing carbon(C) content in the main deuterium(D) plasma. In EAST, the integrated use of discharge cleanings and lithium(Li) coating has positive effects on the suppression of W sputtering source. In the plasma recovery experiments, it is suggested that the W intensity is reduced by approximately 60% with the help of ～35 h Ion Cyclotron Radio Frequency Discharge Cleaning(ICRF-DC) and ～40 g Li coating after vacuum failure. The first wall covered by Li film could be relieved from the bombardment of energetic particles, and the impurity in the vessel would be removed through the particle induced desorption and isotope exchange during the discharge cleanings. In general, the sputtering yield of W would decrease from the source, on the bias of the improvement of wall condition and the mitigation of plasmawall interaction process. It lays important base of the achievement of high-parameter and longpulse plasma operation in EAST. The experiences also would be constructive for us to promote the understanding of relevant physics and basis towards the ITER-like condition.     

Modelling of Edge Plasma on JT-60U Tokamak by Using SOLPS5.0 code

The edge plasma transport code SOLPS5.0 is used for modelling edge plasmas in the experimental shots on JT-60U tokamak and the proles of the radial particle and heat transport coecients D, e and i along the outer midplane have been obtained by tting the code results to the experimental measurement in L-mode shot 39090 and H-mode shots 37851, 37856. The experimental measurement used for tting includes the proles of electron temperature and density along the outer midplane, the pumping speed, the total particle ux from the core boundary to the computational region and the ux density of neutrals near the outer wall. The modelling and tting results show within the pedestal region in H-mode shots 37851 and 37856 the radial particle transport coecient D has larger drop, but, for L-mode shot 39090, the obvious drop of D and e has not been found.     

Effect of Plasma Rotation on Neutral Beam Heating and Current Drive in Tokamaks

For a rapidly rotating plasma, the effects of the resulting Doppler shift have to be included in the neoclassical theory of neutral beam heating, current drive, and plasma transport. In this paper, an improved simulation of neutral beam injection (NBI) and current drive in rotating plasmas is introduced. NBI is simulated using the Monte Carlo code NUBEAM along with the transport code ONETWO. The physical characteristics of heating and current drive for co- and counter-NBI are investigated for non-rotating, co-rotating, and counter-rotating plasmas, all of which can take place in the experiments. In general, it is found that rotation of the plasma can increase the NBI power deposition on the plasma electrons but has little effect on the ions. Moreover, plasma heating by co-NBI is more efficient than that by counter-NBI. For neutral beam current drive, because of the Doppler shift, co-rotation (counter-rotation) of the bulk plasma tends to decrease the co-NBI (counter-NBI) driven current. On the other hand, due to trapping and orbit loss of the fast ions, co-rotation (counter-rotation) has little effect on the counter-NBI (co-NBI) driven current. The results are applied to the forthcoming NBI heating and current drive experiments of the EAST tokamak and should also be useful in the design of experiments in ITER.     

Optimization Study of ICRF Hydrogen Minority Heating in a Deuterium Plasma of EAST

The full wave TORIC code and the Kinetic Fokker-Planck SSFPQL code are combined to perform self-consistent simulations of the ICRF heating in the EAST 2D magnetic configuration.The combined package is applied to the ICRF hydrogen minority heating in a deuterium plasma with the hydrogen concentration up to 10%.The fast wave propagation and absorption properties,power partitions among the plasma species and the RF driven energetic tails have been analyzed.Meanwhile,in order to optimize the ICRF heating,changing the resonance locations has also been considered in EAST plasmas.     

Investigation on impurity transport using LBO technique

1. IntroductionIt is a common phenomenon in the tokamak devices that impurities have a concentrated trend towards the plasma central region. Impurity sourcesand its transport are very complicated, traditionalpassive methods can not be satisfied for the investigation of impurity transport. So transient perturbation methods have been developed and recognizedas the most appropriate for impurity transport measurement.The LBO technique is just the right transient perturbation method for this task…     

First Results of Pellet Injection Experiments on EAST

A new pellet injection system was installed on the EAST tokamak and preliminary experiments were performed during the 2012 run campaign. Typical phenomena associated with deuterium pellet injection into a plasma discharge have been observed including sudden increases of the electron density and H α /D α emission intensity as well as a significant decrease in plasma electron temperature. Profiles have been studied in order to understand the influence of pellet fuelling on EAST discharges. Even though the injector was specifically designed for plasma fu- elling, ELM triggering using the pellet injection has also been tested. In order to find appropriate parameters for triggering ELMs in H-mode plasmas, scanning of the pellet injection speed was employed for pellets injected from both the high field side and low field side of the plasma column. It has been observed that low-speed pellets injected into H-mode plasma from the low-field side could trigger an ELM followed by a number of smaller induced ELMs at about 300 Hz.