Influence of the X-point location on edge plasma transport in the J-TEXT tokamak with a high-field-side single-null divertor

High-density experiments in the high-field-side mid-plane single-null divertor configuration have been performed for the first time on J-TEXT.The experiments show an increase in the highest central channel line-averaged density from 2.73 x 1019 m-3 to 6.49 x 1019 m-3,while the X-point moves away from the target by increasing the divertor coil current.The corresponding Greenwald fraction rises from 0.50 to 0.79.For the impurity transport,the density normalized radiation intensity(absolute extreme ultraviolet and soft x-ray)of the central channel density decreased significantly(＞50％)with an increase in the plasma density.To better understand the underlying physics mechanisms,the 3D edge Monte Carlo code coupled with EIRENE(EMC3-EIRENE)has been implemented for the first time on J-TEXT.The simulation results show good agreement with the experimental findings.As the X-point moves away from the target,the divertor power decay length drops and the scrape-off layer impurity screening effect is enhanced.     

Plasma equilibrium calculation in J-TEXT tokamak

Plasma equilibrium has been calculated using an analytical method. The plasma profiles of the current density, safety factor, pressure and magnetic surface function are obtained. The analytical solution of the Grad–Shafranov(GS) equation is obtained by the variable separation method and compared with the computed results of the equilibrium fitting code EFIT.     

Upgrade of an integrated Langmuir probe system on the closed divertor target plates in the HL-2A tokamak

A newly designed divertor Langmuir probe diagnostic system has been installed in a rare closed divertor of the HL-2A tokamak and steadily operated for the study of divertor physics involved edge-localized mode mitigation, detachment and redistribution of heat flux, etc. Two sets of probe arrays including 274 probe tips were placed at two ports (approximately 180° separated toroidally), and the spatial and temporal resolutions of this measurement system could reach 6 mm and 1 μs, respectively. A novel design of the ceramic isolation ring can ensure reliable electrical insulation property between the graphite tip and the copper substrate plate where plasma impurities and the dust are deposited into the gaps for a long experimental time. Meanwhile, the condition monitoring and mode conversion between single and triple probe of the probe system could be conveniently implemented via a remote-control station. The preliminary experimental result shows that the divertor Langmuir probe system is capable of measuring the high spatiotemporal parameters involved the plasma density, electron temperature, particle flux as well as heat flux during the ELMy H-mode discharges.     

Magnetic configuration flexibility of snowflake divertor for HL-2M

HL-2M (Li, 2013 [1]) is a tokamak device that is under construction. Based on the magnetic coils design of HL-2M, four kinds of divertor configurations are calculated by CORSICA code (Pearlstein et al., 2001 [2]) with the same main plasma parameters, which are standard divertor, exact snowflake divertor, snowflake-plus divertor and snowflake-minus divertor configurations. The potential properties of these divertors are analyzed and presented in this paper: low poloidal field area around X-point, connection length from outside mid-plane to the primary X-point, target plate design and magnetic field shear. The results show that the snowflake configurations not only can reduce the heat load at divertor target plates, but also may improve the magneto-hydrodynamic stability by stronger magnetic shear at the edge. A new divertor configuration, named “tripod divertor”, is designed by adjusting the positions of the two X-points according to plasma parameters and magnetic coils current of HL-2M.     

The timing system on the J-TEXT tokamak

This paper describes the timing system designed to control the operation time-sequence and to generate clocks for various sub-systems on J-TEXT tokamak. The J-TEXT timing system is organized as a distributed system which is connected by a tree-structured optical fiber network. It can generate delayed triggers and gate signals (0 μs–4000 s), while providing reference clocks for other sub-systems. Besides, it provides event handling and timestamping functions. It is integrated into the J-TEXT Control, Data Access and Communication (J-TEXT CODAC) system, and it can be monitored and configured by Experimental Physics and Industrial Control System (EPICS). The configuration of this system including tree-structured network is managed in XML files by dedicated management software. This system has already been deployed on J-TEXT tokamak and it is serving J-TEXT in daily experiments.     

Preliminary study of divertor particle exhaust in the EAST superconducting tokamak

The particle exhaust of the upper tungsten and lower carbon divertors in EAST has been preliminarily studied during the 2016 experimental campaign. The density decay time during terminating gas puffing has been employed as a key parameter to evaluate the divertor particle exhaust performance. Comparative plasma discharges have been carried out on the particle exhaust performance between two toroidal field directions in the upper single null and lower single null divertor configurations. This work has enhanced the understanding of the effects of the in–out asymmetry and divertor geometry on the efficiency of the divertor particle exhaust. In addition, the sensitivity of the particle exhaust capability on different strike point locations has been analyzed. The experimental results are expected to provide important information on the future upgrade of EAST bottom divertor and facilitate the realization of longer pulse operation.     

Design and realization of the J-TEXT tokamak central control system

The Joint Texas Experimental Tokamak (J-TEXT), a medium-sized conventional tokamak, serves as a user experimental facility in the China-USA fusion research community. Development of a flexible and easy-to-use J-TEXT central control system (CCS) is of supreme importance for users to coordinate the experimental scenarios with full integration into the discharge operation. This paper describes in detail the structure and functions of the J-TEXT CCS system as well as the performance in practical implementation. Results obtained from both commissioning and routine operations show that the J-TEXT CCS system can offer a satisfactory and effective control that is reliable and stable. The J-TEXT tokamak achieved high-quality performance in its first-ever experimental campaign with this CCS system.     

Wave-driver options for low aspect ratio steady-state tokamak reactors

Low aspect ratio designs are proposed for steady-state tokamak reactors. Benefits stem from reduced major radius and lessened stresses in the toroidal field coils, resulting in possible cost savings in the tokamak construction. In addition, a low aspect ratio (A=2.6) permits the application of a bundle divertor capable of diverting 3-T fields to a power reactor using STARFIRE technology. Such a low aspect ratio is possible with the elimination of poloidal field coils in the central hole of the tokamak, which implies a need for noninductive current drive. Several plasma waves are considered for this application, and it appears likely that a candidate can be found which reduces the electric power for current maintenance to an acceptable value.     

Soft landing of runaway currents by ohmic field in J-TEXT tokamak

Recently, experimental studies on the soft landing of RE current by ohmic (OH) field have been performed in J-TEXT tokamak, as a possible auxiliary method to dissipate the RE current. With optimized horizontal displacement control of the RE beam, the toroidal electric field has been scanned from 1.6 to - 0.3 V m ^-1 during the RE plateau phase. The growth rate of RE currents and the evolution of hard x-ray (HXR) emissions have been studied. It is found that when the toroidal electric field is less than 7–12 times the theoretical critical electric field, the decay of REs could be achieved. The dissipation rate by the ohmic field can reach a maximum value of 3 MA s^-1.Furthermore, the results of HXR spectra analysis indicate the different behaviors of HXR emissions under the condition of different toroidal electric fields. The analysis of the ionized argon emissions and magnetic fluctuations shows that under the condition of different toroidal electric fields, the physical process of RE generation may be different.     

Achieving temporary divertor plasma detachment with MARFE events by pellet injection in the EAST superconducting tokamak

A new pellet injection system has been equipped on the experimental advanced superconducting tokamak(EAST) in the 2012 campaign,with a pellet size of Ф 2 mm?×?2 mm,a frequency of1 Hz–10 Hz and velocity of 150 m s~(-1)–300 m s~(-1).The deuterium pellet is well-known for plasma fuelling as well as for? triggering the edge localized mode(ELM).In the 2012 campaign,pellet injection experiments were successfully carried out on EAST.Temporary plasma detachment achieved by deuterium pellets has been observed in a double null(DN) divertor configuration,with multi-pellet injections at a repetition frequency of 2 Hz.The partial detachment of the outer divertors and complete detachment of the inner divertors was achieved after 35 ms of each pellet injection,which have a duration of 30–60 ms with the maximum degree of detachment(DOD) reaching 3.5 and 37,respectively.Meanwhile,the multifaceted asymmetric radiation from the edge(MARFE) phenomena was also observed at the high field side(HFS) near both the lower and upper X-points with radiation loss suddenly increased to about 15%–70%,which may be the main cause of divertor plasma detachment.The temporary detachment induced by pellet injection may act as a new way to study divertor detachment behaviors.     

Space-resolved vacuum-ultraviolet spectroscopy for measuring impurity emission from divertor region of EAST tokamak

The measurement of impurity distribution in the divertor region of tokamaks is key to studying edge impurity transport. Therefore, a space-resolved vacuum-ultraviolet (VUV) spectrometer is designed to measure impurity emission in the divertor region on EAST. For good spectral resolution, an eagle-type VUV spectrometer with 1 m long focal length with spherical holograph grating is used in the system. For light collection, a collimating mirror is installed between the EAST plasma and the VUV spectrometer to extend the observing range to cover the upper divertor region. Two types of detectors, i.e. a back-illuminated charge-coupled device detector and a photomultiplier-tube detector, are adopted for the spectral measurement and high-frequency intensity measurement for feedback control, respectively. The angle between the entrance and exit optical axis is fixed at 15°. The detector can be moved along the exit axis to maintain a good focusing position when the wavelength is scanned by rotating the grating. The profile of impurity emissions is projected through the space-resolved slit, which is set horizontally. The spectrometer is equipped with two gratings with 2400 grooves/mm and 2160 grooves/mm, respectively. The overall aberration of the system is reduced by accurate detector positioning. As a result, the total spectral broadening can be reduced to about 0.013 nm. The simulated performance of the system is found to satisfy the requirement of measurement of impurity emissions from the divertor area of the EAST tokamak.     

Machine learning application to predict the electron temperature on the J-TEXT tokamak

The reliability of diagnostic systems in tokamak plasma is of great significance for physics researches or fusion reactor. When some diagnostics fail to detect information about the plasma status, such as electron temperature, they can also be obtained by another method: fitted by other diagnostic signals through machine learning. The paper herein is based on a machine learning method to predict electron temperature, in case the diagnostic systems fail to detect plasma temperature. The fully-connected neural network, utilizing back propagation with two hidden layers, is utilized to estimate plasma electron temperature approximately on the J-TEXT. The input parameters consist of soft x-ray emission intensity, electron density, plasma current, loop voltage, and toroidal magnetic field, while the targets are signals of electron temperature from electron cyclotron emission and x-ray imaging crystal spectrometer. Therefore, the temperature profile is reconstructed by other diagnostic signals, and the average errors are within 5%. In addition, generalized regression neural network can also achieve this function to estimate the temperature profile with similar accuracy. Predicting electron temperature by neural network reveals that machine learning can be used as backup means for plasma information so as to enhance the reliability of diagnostics.     

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.     

Toroidal component of velocity for geodesic acoustic modes in the edge plasmas of the J-TEXT tokamak

The toroidal component of the velocity for geodesic acoustic modes (GAMs) is first demonstrated.Multiple Langmuir probe arrays set up near the top tokamak of the J-TEXT were utilized for this study.A significant peak at the GAM frequency is observed in Mach number fluctuations.The toroidal velocity for the GAMs is estimated as ～10-100 m s-1 and increases with the poloidal velocity.The ratio of toroidal component to the poloidal one of the velocity is mainly located in the interval between 0.3 and 1.0.With higher safety factors q,the ratio almost does not change with decreasing the safety factor,whereas it goes up sharply at low q.The coherencies between poloidal electric fields and Mach number fluctuations in turbulence frequency bands are also evaluated,and are higher than those between radial electric fields and Mach number fluctuations.     

Preliminary 2D code simulation of the quasi-snowflake divertor configuration in the FAST tokamak

The behaviour of the SOL in FAST for the new quasi snowflake configuration of the divertor is compared with that of the conventional single null case for three main scenarios: reference, advanced and extreme H-mode. The flexible, quick and versatile 2D code TECXY is used. The main physics processes occurring at the edge are carefully taken into account but the neutral dynamics is simplified. Even though the status close to detachment conditions cannot be detailed, new phenomena generated by the configuration are clearly highlighted. The heat load is strongly mitigated to a level that is much easier to tackle with the present technology. Mitigation is always stronger than expected from the magnetic topology only, especially in the high density regimes. This is attributed to the much longer time that the SOL particles spend in the quite cold region of the X point where mostly the connection length of the magnetic lines is strongly increased and hence to the much more intense interaction of the plasma particle with the background neutrals.     

Improving concept design of divertor support system for FAST tokamak using TRIZ theory and AHP approach

The paper focuses on the application of the Theory of Inventive Problem Solving (TRIZ) to divertor Remote Handling (RH) issues in Fusion Advanced Studies Torus (FAST), a satellite tokamak acting as a test bed for the study and the development of innovative technologies oriented to ITER and DEMO programs. The objective of this study consists in generating concepts or solutions able to overcome design and technical weak points in the current maintenance procedure. Two different concepts are designed with the help of a parametric CAD software, CATIA V5, using a top-down modeling approach; kinematic simulations of the remote handling system are performed using Digital Mock-Up (DMU) capabilities of the software. The evaluation of the concepts is carried out involving a group of experts in a participative design approach using virtual reality, classifying the concepts with the help of the Analytical Hierarchy Process (AHP).     

A study on nuclear analysis of the divertor region of the CFETR

This paper presents the nuclear analysis performance of the Chinese Fusion Engineering Test Reactor(CFETR)divertor region using the MCNP-5 Monte Carlo N-particles code in a 3D geometry model.We assessed the nuclear responses of the divertor region component systems and evaluated their shielding capability,which can support the development strategy of the physical and engineering design of the CFETR.Model specification based on the latest CAD model of the CFETR divertor has been integrated into the CFETR MCNP reference model with a major/minor radius R=7.2 m/a=2.2 m in the 22.5° model,and a fusion-power range of around 1-1.5 GW.The nuclear heating and radiation damage of the divertor system are enhanced compared to that of the ITER and the earlier CFETR design.The initial nuclear responses of the toroidal field coil and vacuum vessel systems showed that the shielding of the current divertor design is not sufficient and optimization work has been carried out.We also carried out calculations and analysis using a hypothetical operating scenario of over 14 years.An excellent improvement in the nuclear performance has been obtained by the improved additional shielding block in the divertor region when referring to the ITER design limit,which can support the design of the future update of the divertor region systems of the CFETR.     

Acceleration optimization of real-time equilibrium reconstruction for HL-2A tokamak discharge control

Real-time equilibrium reconstruction is crucially important for plasma shape control in the process of tokamak plasma discharge.However,as the reconstruction algorithm is computationally intensive,it is very difficult to improve its accuracy and reduce the computation time,and some optimizations need to be done.This article describes the three most important aspects of this optimization:(1) compiler optimization;(2) some optimization for middle-scale matrix multiplication on the graphic processing unit and an algorithm which can solve the block tri-diagonal linear system efficiently in parallel;(3) a new algorithm to locate the XO point on the central processing unit.A static test proves the correctness and a dynamic test proves the feasibility of using the new code for real-time reconstruction with 129?×?129 grids;it can complete one iteration around 575 μs for each equilibrium reconstruction.The plasma displacements from real-time equilibrium reconstruction are compared with the experimental measurements,and the calculated results are consistent with the measured ones,which can be used as a reference for the real-time control of HL-2 A discharge.     

A brief review of the development and optimization of the three-dimensional magnetic configuration system in the J-TEXT tokamak

The three-dimensional (3D) magnetic configuration system in the J-TEXT tokamak has featured in many experimental studies. The system mainly consists of three subsystems: the static resonant magnetic perturbation (SRMP) system, the dynamic resonant magnetic perturbation (DRMP) system and the helical coil system. The SRMP coil system consist of two kinds of coils, i.e. three six-loop coils and two five-loop coils. It can suppress tearing modes with a moderate strength, and may also cause mode locking with larger amplitude. The DRMP coil system consists of 12 single-turn saddle coils (DRMP1) and 12 double-turn saddle coils (DRMP2). Its magnetic field can be rotated at a few kHz, leading to either acceleration or deceleration of the tearing mode velocity and the plasma rotation. The helical coil system consists of two closed coils, and is currently under construction to provide external rotational transform in J-TEXT. The 3D magnetic configuration system can suppress tearing modes, preventing and avoiding the occurrence of major disruption.     

Current drive operation in a tokamak

Three modes of current drive operation in a tokamak — continuous, cyclic, and rfinitiated-are studied for air core and iron core transformer. It is found that the air core transformer is in general more flexible than the iron core transformer for current drive operation. For continuous operation, the shutoff time of the Ohmic heating circuit of the air core transformer can be reduced to zero by using a bias current. On the other hand, the shutoff time of the iron core transformer remains finite even if the bias current is used, because of hysteresis. For cyclic operation, methods of shortening the recharging time are investigated for both types of transformer. The effects of the transformer on rf-initiated operation are investigated. A model design of a saturable iron core tokamak for current drive experiments is also presented.