Simulation of Low–High Transition by Neutral Beam Injection in Edge Plasma of Small Size Divertor Tokamak

This paper reports simulation of L–H transition by fluid transport code B2SOLPS0.5.2D at low ion plasma density on neutral beam injection (NBI) in the edge plasma of small size divertor tokamak. The simulation provides the following results: (1) the transition is possible at plasma density 2 × 10¹⁹ m^?3 with NBI at temperature heating T^heating 3.62 keV. (2) The simulation predicts the generation of large negative radial electric field E _r, which is thought to help L–H transition during NBI, is suggested in the edge plasma of small size divertor tokamak. (3) The toroidal current density in the edge plasma of small size divertor tokamak is plasma density and direction of NBI dependence. (4) Parallel flux transport by anomalous viscosity (turbulent) through separatrix leads to the variation of toroidal current density.     

Experiment Results about the Relationship of Edge Turbulence and Atomic Processes in the HT-7 Tokamak

Using a reciprocating Langmuir probe system, the boundary plasma behaviors were investigated before and after lithium/silicon coating. Accompanying the effective reduction of impurity radiation, strong shears of radial electric field and poloidal velocity came into being and the turbulence suppression and de-correlation were observed in the edge region of coated wall plasma. All these led to the reduction of the edge transport and improvement of plasma confinement. In the central line averaged density scanning experiments, an enhanced shear of the radial electric field was observed in the edge plasma with the increase of the density, which may account for the enhancement of the transport barrier and the improvement of particle confinement.The results suggest a close link between wall conditions and boundary plasma. In addition to the relationship, (～Te)/Te ～(～n)n/ne and θ_(～T)e(～n)e ～π, had been observed in the plasma edge region, which indicates the important role of the ionization and radiation in turbulence driving.     

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%。     

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.     

Improvement of Plasma Performance Using Carbon Pellet Injection in Large Helical Device

A cylindrical carbon pellet with a size of 1.2L?1.2? to 1.8L?1.8?mm and a velocity of 100 to 300 m/s was injected into Large Helical Device (LHD) for an efficient fueling based on its deeper deposition instead of hydrogen gas puffing and ice pellet injection. Electron density increment of ?ne=1014cm-3 is successfully obtained by single carbon pellet injection without plasma collapse. Typical density and temperature of the ablation plasma of the carbon pellet, e.g., 6.5x1016cm-3 and 2.5eV for CII, are examined respectively by spectroscopic method. A confinement improvement up to 50% compared to ISS-95 stellarator scaling is clearly observed in a relatively low-density regime of ne=2 to 4?1013cm-3, and high ion temperature Ti(0) of about 6keV is also observed with an internal transport barrier at ne=1.2?1013cm-3. In particular, the improvement in the ion temperature largely exceeds that observed in hydrogen gas- puffed discharges, which typically ranges below 3 keV.     

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.     

A Study of the von Neumann Stability Analysis of a Semi-implicit Numerical Method Applied Over the Radial Impurity Transport Equation in Tokamak Plasma

The radial impurity transport equation for tokamak plasma is a form of diffusion–convection–reaction equation. The impurity transport equation is solved to determine the distribution of impurity (non-fuel) ion species with different ionization states perpendicular to magnetic surfaces of tokamak plasma. The equation for each charge (ionization) state Z is a non-linear, second-order in space, first-order in time, parabolic partial differential equation coupled to the previous Z???1 and the next Z?+?1 charge states of the impurity species through its reaction term. The number of differential equations to be solved simultaneously is hence determined by the number of ionization states of the impurity species studied. The solution to the set of these coupled equations can be obtained using a semi-implicit numerical method applied on it. The present study describes the application of von Neumann stability analysis over the semi-implicit numerical method applied over the radial impurity transport equation and determines a generic stability criterion for the method. The stability analysis is further illustrated using the geometry of Aditya tokamak installed at the Institute for Plasma Research Gandhinagar, India as an example. The impurity species considered is oxygen (Atomic number?=?8). This leads to a set of eight coupled equations for charge states Z?=?1 to 8 over which von Neumann analysis is illustrated in present study.     

Effect of edge magnetic island on carbon screening in the J-TEXT tokamak

The effect of externally applied resonant magnetic perturbation(RMP)on carbon impurity behavior is investigated in the J-TEXT tokamak.It is found that the m/n=3/1 islands have an impurity screening effect,which becomes obvious while the edge magnetic island is generated via RMP field penetration.The impurity screening effect shows a dependence on the RMP phase with the field penetration,which is strongest if the O point of the magnetic island is near the low-field-side(LFS)limiter plate.By combining a methane injection experimental study and STRAHL impurity transport analysis,we found that the variation of the impurity transport dominates the impurity screening effect.The impurity diffusion at the inner plasma region(r/a＜0.8)is enhanced with a significant increase in outward convection velocity at the edge region in the case of the magnetic island's O point near the LFS limiter plate.The impurity transport coefficient varies by a much lower level for the case with the magnetic island's X point near the LFS limiter plate.The interaction of the magnetic island and the LFS limiter plate is thought to contribute to the impurity transport variation with the dependence on the RMP phase.A possible reason is the interaction between the magnetic island and the LFS limiter.     

Radiation blistering of Nb Implanted sequentially with helium ions of different energies (3–500 keV)

Cold rolled, polycrystalline niobium samples were irradiated at room temperature with ⁴He⁺ ions sequentially at 14 different energies over an energy range from 3–500 keV. The dose for each energy was chosen to give an approximately uniform concentration of helium between the implant depths corresponding to 3 keV and 500 keV. In one set of experiments the irradiations were started at the Kurchatov Institute with 3-keV ⁴He⁺ ions and extended up to 80 keV in several steps. Subsequently, the same target area was irradiated with ⁴ He⁺ ions at Argonne National Laboratory (ANL) starting at 100 keV and increased to 500 keV in steps of 50 keV. Another set of irradiations was started at ANL with 500-keV ⁴ He⁺ ions and continued with ion energies decreasing to 100 keV. Subsequently, the same area was irradiated at the Kurchatov Institut starting at 80 keV and continued with ion energies decreasing to 3 keV. Both sets of irradiations were completed for two different total doses, 0.5 C cm^?2 and 1.0 C cm^?2.     

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.     

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.     

Heliotron E Results and Large Helical Project

Heliotron E(H-E) experiment was started in 1980. Until 1987 high power heating experiments for improving plasma parameters have almost finished. H-E firstly demonstrated that ECR heated plasmas are usable for target plasmas of NBI or ICRF heating to obtain high density and high temperature currentless plasmas. The highest electron temperature is 1.5keV and ion temperature is 1.6keV and both are realized in the low density regime of <n> (average density) ≤10¹³cm^?3.

H-E also showed that the currentless plasmas have no major disruption and quasi-steady plasmas are confined with controlling impurity ions by titanium gettering and carbon coating.

H-E also obtained <β> (average β) –2%, which is the highest value realized in helical systems, with <n–8×l0¹³cm^?3 and Te(0)–T_i(0)–350 eV at B₀ (magnetic field at the magnetic axis) =0.94 T. In the high β experiments pressure-driven instabilities were observed for peaked pressure profiles and sometimes relaxation oscillations similar to the tokamak internal disruptions were observed.

In the ECRH plasmas neoclassical transport is dominant in the region inside the half radius. However, global confinement time τ_E follows the scaling law τ_E ∝<n>^0.66P_heat ^?0.53 which is different from the neoclassical scaling law. Here P_heat denotes the net heating power.

Based on the H-E results, a new large helical system design study has started in 1986. The plasma parameters entering the regime of <n>τ_E<T> (2–3)× 10¹⁹m^?3?S?keV is investigated, which is about one tenth of fusion plasma condition. From the transport code studies and empirical scaling law based on the H-E results, R=(4×5)m, ā=(50–60)cm and B_o=4T are required to satisfy the above condition with P_heat=20MW. The design study to fix the magnetic field configuration is progressing. Expected one is l=2 and m=10 with additional poloidal coils, where m is a toroidal period number. The magnetic field is produced by superconducting coil and long pulse operation will be tested, if continuous heating is available.     

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.     

Turbulent Transport in the Tokamak Edge Plasma and SOL Region

In this work, the turbulent transport in the edge plasma and Scrape-Off Layer (SOL) region of IR-T1 tokamak at the presence of biased limiter has been investigated and analyzed. The time and radial evolution of floating potential, electric field and turbulent transport have been measured by using two arrays of the Langmuir probes in both the radial and poloidal directions. The analyses have been done by the Fast Fourier Transport method and spectral features of them are obtained with the help of the standard Auto-Correlation technique and modified covariance power spectral density estimate. The probability distribution function and actual transfer function magnitude of the radial and poloidal turbulent transport (Γ_r and Γ_p) have been investigated and compared in the edge plasma and SOL region. Also the histogram of turbulent transport has been analyzed and compared in the edge and SOL at presence of positive limiter biasing. The results show that in the edge plasma poloidal turbulent transport (Γ_p) is about of 60 % more than SOL region whereas radial turbulent transport (Γ_r) is about of 40 % less. During the application of positive biasing, it was found that Γ_r in the IR-T1 reduces by about 80 % in the edge plasma and 45–50 % in the SOL. Increase of Γ_p is about of 50 % after applied positive biasing in the edge while it increases 70 % nearly, in the SOL. Consequently, the improvement in confinement can be obtained for positive limiter biasing.     

Pb离子辐照注碳a-SiO2薄膜的光致发光性能

用湿氧化法在单晶硅表面生长了非晶态SiO2薄膜,进行120 keV C离子注入和950 MeV Pb离子辐照,用荧光光谱分析样品发光特性的改变.结果发现,C离子注入和高能Pb离子辐照均能显著影响样品的发光特性,且荧光光谱的改变强烈依赖于注入和辐照剂量,预示不同注入和辐照剂量将导致不同的发光结构形成.对注入和辐照造成薄膜...     

Internal Transport Barrier in Edge Plasma of Small Size Divertor Tokamak Using Neutral Beam Injection

We model the internal transport barrier “ITB” in edge plasma of small size divertor tokamak with B2SOLPS0.5.2D fluid transport code. The simulation results demonstrated the following: (1) we control the internal transport barrier by altering the edge particle transport through changes the edge toroidal rotation which agree with the result of Burrell et al. (Edge Pedestal control in quiescent H-mode discharges in DIII-D using co-plus counter-neutral beam injection, Nucl Fusion, 49, 085024 (9pp) in 2009). (2) The radial electric field has neoclassical nature near separatrix with discharge by co-injection NBI. (3) The toroidal plasma viscosity has strong influence on the toroidal velocity.     

Radiative Cooling Rates for Low-Z Impurities in Non-coronal Equilibrium State

The local or transient radiation losses in tokamak plasmas can greatly exceed those in the coronal equilibrium. This excess is especially pronounced at the plasma edge. The reason for the increase of radiation in a peripheral plasma is as follows. The impurities are lost fast from the plasma edge and the new impurity source is supplied to this region. The charged states of impurities, therefore, do not reach their coronal equilibrium ones. These impurity ions have more electrons than those in the coronal equilibrium, and as a result emit the higher radiation power. In the simplest case, the non-coronal radiative rate can be determined only by two parameters: the electron temperature \(T_{\text {e}}\) and the so-called “residence parameter” \(n_{\text {e}}\tau _{\text {i}}\), where \(\tau _{\text {i}}\) is the impurity residence time in the plasma. Despite the strong simplification, such an approach allows to do simple estimates of non-coronal radiation. In this paper, two dimensional polynomial fits describing radiative cooling rates and mean charge are obtained for eight impurity species: helium, lithium, beryllium, carbon, nitrogen, oxygen, neon, and argon. The results are presented in figures and tables. The figures show curves calculated from the original atomic database and least-squares polynomial fits to these curves. The tables contains coefficients for this fits. The obtained fits can be useful for qualitative estimates and simple numerical calculations.     

蒙特卡罗模拟确定γ射线衰减系数函数及参数

在中低密度样品中,γ射线的线衰减系数主要由γ射线能量和样品密度决定,采用MCNP(Monte Carlo N Particle Transport Code)程序模拟计算了多种γ射线能量和多种样品密度条件下的线衰减系数,对线衰减系数模拟值进行多元非线性回归,确定了以γ射线能量和样品密度为因变量的线衰减系数函数及参数。实验测定了三种能量γ射线在6种不同密度样品中的线衰减系数值,并与模拟所得函数值进行比较分析。结果表明,所得函数值与实验值的相对误差均在7%以内,蒙特卡罗程序计算所得函数值与实验测量值较为吻合,所采用的函数模型准确验证了线衰减系数与γ射线能量、样品密度之间的关系特征。     

Determination of ion fraction and energy analysis of sputtered particles from deuterium bombarded surfaces

Ion energy distributions and sputtering yields have been measured in the presence of various background gases for deuterium, helium and argon ions in the 1–40 keV energy range. Alteration of the surface chemistry of hydride forming metals such as Ti, Y and V by exposure to hydrogen has a significant effect on the charge state of the sputtered particles. Under conditions likely to prevail in plasma devices and possible reactors, the ion/neutral fraction may be drastically increased, reaching ~41% for Ti. Such large ion fractions for metal targets have been previously observed only by sputtering with oxygen ions and present the possibility of improved impurity control in plasma devices. Both gas adsorption and recoil implantation are involved in the mechanism determining the ion fraction of sputtered products.     

Formation of Electron Internal Transport Barrier in Edge Plasma of Small Size Divertor Tokamak

The formation of electron internal transport barrier (EITB) during using counter-neutral beam injection (NBI) heating in the edge plasma of small size divertor tokamak can be simulated by using fluid transport code B2SOLPS0.5.2D. The results of simulations give us the following: (1) Plasma heating with counter-neutral beam injection leads to, strong, parabola type electron internal transport barrier (EITB) was formed in the edge plasma of small size divertor tokamak. (2) In case of plasma heating by counter-neutral beam injection, the radial electric field shear (E _r –gradient) was increased, while electron transport coefficients were reduced in conjunction with the formation of electron internal transport barrier (EITB). (3) The plasma heating by counter-neutral beam injection play significantly role in redistribution of parallel (toroidal) velocity in edge plasma of small size divertor tokamak.