Experimental Advanced Superconducting Tokamak Plasma Research Articles

An E-band multi-channel Doppler backscattering system on EAST.

An E-band (60-90 GHz) multi-channel Doppler backscattering (DBS) system with X-mode polarization has been installed on the Experimental Advanced Superconducting Tokamak (EAST), which can measure the turbulence at five different radial locations simultaneously. This system can launch 31 fixed microwave frequencies in the range of 60-90 GHz with a 1GHz interval into the plasma, and five probing signals are selected by employing a reference signal and multiple filters. During experiments, the frequency of the reference signal is tunable in the E-band, and the selected probing signals can be changed as needed without any other adjustments, which can be performed in one shot or between shots. Furthermore, the incident angle can be adjusted from -10° to 20°, and the wavenumber range is 4-25cm-1 with a wavenumber resolution of Δk/k ≤ 0.35. Ray tracing simulations are employed to calculate the scattering locations and the perpendicular wavenumber. In this article, the hardware design, ray tracing, and initial results obtained from the EAST plasma will be presented.

Mechanical effects: challenges for high-field superconducting magnets.

Due to its clean products and sufficient raw materials, fusion energy is expected to become one of the main solutions of the energy crisis and ensuring the sustainable development of human society, which is a long-term strategic frontier field. The promise of fusion energy is to constrain the motion of high-temperature plasma by the high magnetic field generated by superconducting magnets, and then achieve controllable thermonuclear fusion. Fusion power is proportional to the fourth power of the magnetic field strength. Thus, future commercial fusion reactors need a higher magnetic field as the basis for sustainable development [1]. In order to verify the scientific and technological feasibility of fusion energy, China, the United States, the European Union, Russia et al. have jointly participated in the construction of the International Thermonuclear Fusion Test Reactor (ITER), which is expected to produce the first plasma discharge by 2025 [2]. Currently, China is leading the world in many fields of fusion energy research. For example, the experimental advanced superconducting Tokamak (EAST) whole-superconducting Tokamak located at the Institute of Plasma Physics in the Chinese Academy of Sciences has achieved a repeatable world record of stable plasma operation at 120 million degrees Celsius for 101 seconds, which provides a solid foundation for ITER and also China's future Independent Building Fusion Reactor (https://www.cas.cn/syky/202105/t20210528_4790357.shtml). Prof. Jiangang Li, an academician of the Chinese Academy of Engineering, participated in and completed the design and construction of EAST plasma facing componments (PFCs) engineering by the support of the national '9th five-year plan' major scientific and technological infrastructure, and presided over the completion of the national '11th five-year plan' major scientific and technological infrastructure-EAST auxiliary heating system project. He also hosted the national '13th five-year plan' major scientific and technological infrastructure-Integrated Research Facility for Critical Systems of fusion reactor comprehensive research facility for fusion technology (CRAFT). Many important scientific and technological problems have been solved and overcome by Prof. Li and his co-workers, which puts China's plasma physics research and fusion engineering technology at the forefront of global engineering.

Events detection and handling based on hierarchical finite state machines in EAST PCS

Experimental Advanced Superconducting Tokamak (EAST) is committed to long pulse high performance operation where plasma parameters will necessarily approach stability limits, which sets a higher demand on real-time events detection and handling (EDH). In order to enhance the current EAST plasma control system (PCS) events handling ability and explore a new intelligent scheduling mode for China Fusion Engineering Testing Reactor (CFTER) PCS, the EDH algorithm based on hierarchical finite state machines (HFSM) is designed in EAST plasma control system (PCS). The events and states in PCS were designed, and we successfully validated the state transition and algorithm scheduling. Moreover, a new event detection and processing for controlling hot spots of in-vessel components from hot spots is developed, and a simulation experiment is carried out in the closed-loop test environment between PCS and MATLAB/Simulink. The simulation results show that the EDH algorithm could capture the exception in time and implement effective strategies with the change of hot spots temperature. The temperature of hot spots were prevented from reaching a drastic situation, which proved the effectiveness of the EDH to protect the internal components of the device.

Application of local Monte Carlo method in neutronics calculation of EAST radial neutron camera

The Local Monte Carlo (LMC) method is used to solve the problems of deep penetration and long time in the neutronics calculation of the radial neutron camera (RNC) diagnostic system on the experimental advanced superconducting tokamak (EAST), and the radiation distribution of the RNC and the neutron flux at the detector positions of each channel are obtained. Compared with the results calculated by the global variance reduction method, it is shown that the LMC calculation is reliable within the reasonable error range. The calculation process of LMC is analyzed in detail, and the transport process of radiation particles is simulated in two steps. In the first step, an integrated neutronics model considering the complex window environment and a neutron source model based on EAST plasma shape are used to support the calculation. The particle information on the equivalent surface is analyzed to evaluate the rationality of settings of equivalent surface source and boundary. Based on the characteristic that only a local geometric model is needed in the second step, it is shown that the LMC is an advantageous calculation method for the nuclear shielding design of tokamak diagnostic systems.

Observation of direct power deposition of ICRF in EAST plasma boundary

Waves in the ion-cyclotron range of frequencies (ICRF) used for plasma heating are known to interact strongly with the scrape-off layer (SOL) plasma, potentially limiting ICRF performance. We present the evidence of direct power deposition of ICRF in the Experimental Advanced Superconducting Tokamak (EAST) plasma boundary, with two ICRF antennas and 12 MW of ICRF source power in support of high-power and long-pulse operation. By applying Retarding Field Analyzer (RFA) diagnostic, a sharp rise in average collected ion parallel energy was obtained during the period of ICRF heating as the RFA probe stayed still at 2 mm behind the main limiter which indicates that a certain amount of power is deposited near the limiter instead of being absorbed by the main plasma. A significant increase of the target floating potential in electron temperature measured by Divertor Langmuir Probes arrays was observed. The wall temperature in the far SOL was also enhanced. As a consequence, an enhancement of the carbon and tungsten impurity concentration was observed both at the plasma boundary and core. Besides, plasma radiation, the impurity concentration of the boundary and core increased, and the temperature rise of the first wall was also observed, which are intuitive and powerful pieces of evidence of ICRF direct power deposition at the boundary.

Development of High-Speed Image Acquisition and Processing System for Real-Time Plasma Control on EAST

In order to realize advanced plasma control based on visible cameras, a high-speed image acquisition and processing system has been developed recently on Experimental Advanced Superconducting Tokamak (EAST). This system is optimized in many ways to achieve high-speed, real-time, low-latency performance, and to load multiple acquisition cards simultaneously. The acquisition rate of this system can be close to 10 000 FPS when the frame size is set as 320 <inline-formula xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"> <tex-math notation="LaTeX">$\times240$ </tex-math></inline-formula> and the pixel depth is set as 8 bits. DMA is used for high-speed data transmission; the memory copy function is optimized for reducing the time cost on data memory reading and writing. Besides, the visualization subsystem based on the Python web can communicate with the acquisition machines and also can synthesize data of multiple acquisition machines in real time to perform image fusion and access display. In addition, a thermal event recognition function based on visible imaging is also included in this system. The convolutional neural network (CNN) model of hot spots and multifaceted asymmetric radiation from the edge (MARFE) detection for EAST plasma discharge has been developed so far. The detection results can be visualized in quasi-real time. In terms of data storage, a new data storage format is designed; a GUI for off-line data analysis and processing based on MATLAB is provided.

Study of microscale surface morphology changing of the retroreflector in EAST by means of laser speckle interferometry technique

• The surface morphology of the retroreflector served in EAST is first time reconstructed through a 3D visual exhibition. • Three internal reflection planes of the retroreflector served in EAST appear severely different surface morphology features. • Laser speckle interferometry technique is a reliable surface morphology characterization approach for PFMs. The retroreflectors in EAST (Experimental Advanced Superconducting Tokamak) are important optical components of Polarimeter-Interferometer (POINT) system, working as the first mirrors to reflect the incident far-infrared laser. The surface morphology changing of the retroreflector has a direct impact on the optical reflection accuracy and it will generate a significant influence on EAST plasma density measurement and plasma current distribution diagnosis. Therefore, it is essential to research the surface morphology changing of the retroreflector systematically. Laser speckle interferometry technique is considered as a most promising method for real-time monitoring the PFMs surface morphology changing at plasma discharge intervals in the near future and its measurement accuracy and reliability have been adequately verified in the laboratory. One retroreflector served in 2019–2020 EAST Campaign is selected as a representative, compared with a brand-new retroreflector, to demonstrate that microscale surface morphology changing occurred on the internal reflection planes after bearing a whole EAST campaign plasma interaction. Meanwhile, the surface morphology of each internal reflection plane appears severely different features. Further research and analysis of the surface morphology are more contributed to evaluating the damage states of the retroreflector and providing more suggestions for the improvement and protection of the retroreflector. The surface morphology reconstruction results are given through a 3D visual exhibition and laser speckle interferometry technique is proposing a new research perspective for PWI research.

Development of a wide-range neutron flux monitoring system in EAST

A wide measurement range for a neutron flux monitoring (NFM) system is developed in the Experimental Advanced Superconducting Tokamak (EAST) for fusion neutron yield measurement, comprising an 3He proportional counter, a 235U fission chamber (FC), and a wide dynamic range signal-processing circuit. Two high-sensitivity 3He detectors operate in counting mode, which is suitable for low neutron yield (Yn) conditions with ohmic and radio-frequency heating of plasma. Two other FCs are adapted with newly developed digitizers, operating in counting and campbelling modes to satisfy the conditions of wide dynamic range measurement required by neutral beam injection (NBI) experiment. The anti-magnetic-field interference of the FC is examined using two permanent magnets, whose efficiency are calibrated with 2.46 MeV neutrons produced by a Van de Graaff accelerator. Furthermore, the NFM system is implemented in-situ calibration with a 252Cf neutron source and a remote handling. As a routine diagnostic, the NFM system is successful obtaining the fusion neutron yield for ion cyclotron resonance heating (ICRH) and NBI heating performance evaluation in EAST plasma experiment.

Development of the quasi-optical combiner systems for density profile reflectometers on the EAST tokamak

The reflectometry diagnostics on experimental advanced superconducting tokamak (EAST) are composed of Q-band (32–56 GHz), V-band (48–76 GHz) and W-band (72–110 GHz) microwave reflectometers with X-mode polarization. The three reflectometers with separate frequency bands were launched independently in previous experiments and now are combined together by applying advanced microwave technology. A quasi-optical (QO) combiner/de-combiner by using frequency selective surfaces (FSSs) has been built to combine the three bands so that these waves (32–110 GHz) can be transmitted to the antenna by using one single oversized waveguide. Two double-ridged horns are used for launcher and receiver. The received waves are decoupled by using a QO de-combiner, the same with the combiner. The designed parameters and laboratory test results of the upgraded system have been presented. The density pedestal evolution during the edge localized modes (ELMs) in EAST plasma have been presented as an example for the application of the updated reflectometers.

R&D progress of high power ion source on EAST-NBI

The neutral beam injector (NBI) system was designed and developed mainly for the plasma heating on the Experimental Advanced Superconducting Tokamak (EAST). The high power ion source is the key part of the NBI. A hot cathode ion source was used on the EAST-NBI. The ion source was conditioned on the ion source test bed with hydrogen gas and achieved the designed parameters. The deuterium gas was used when it moved to the EAST-NBI. The main performance of the ion source on EAST is presented in this paper. The highest beam power of 4.5 MW in NBI-1 and 2.75 MW in NBI-2 was achieved. The total neutral beam power is about 4.5 MW. The long pulse beam of 100 s is injected into the EAST plasma too.

Overview of decade-long development of plasma-facing components at ASIPP

The first EAST (Experimental Advanced Superconducting Tokamak) plasma ignited in 2006 with non-actively cooled steel plates as the plasma-facing materials and components (PFMCs) which were then upgraded into full graphite tiles bolted onto water-cooled copper heat sinks in 2008. The first wall was changed further into molybdenum alloy in 2012, while keeping the graphite for both the upper and lower divertors. With the rapid increase in heating and current driving power in EAST, the W/Cu divertor project was launched around the end of 2012, aiming at achieving actively cooled full W/Cu-PFCs for the upper divertor, with heat removal capability up to 10 MW m−2. The W/Cu upper divertor was finished in the spring of 2014, consisting of 80 cassette bodies toroidally assembled. Commissioning of the EAST upper W/Cu divertor in 2014 was unsatisfactory and then several practical measures were implemented to improve the design, welding quality and reliability, which helped us achieve successful commissioning in the 2015 Spring Campaign. In collaboration with the IO and CEA teams, we have demonstrated our technological capability to remove heat loads of 5000 cycles at 10 MW m−2 and 1000 cycles at 20 MW m−2 for the small scale monoblock mockups, and surprisingly over 300 cycles at 20 MW m−2 for the flat-tile ones. The experience and lessons we learned from batch production and commissioning are undoubtedly valuable for ITER (International Thermonuclear Experimental Reactor) engineering validation and tungsten-related plasma physics.

Retarding field analyzer for the EAST plasma boundary.

A novel bi-directional Retarding Field Analyzer (RFA) probe has been installed on a fast reciprocating drive system on the Experimental Advanced Superconducting Tokamak (EAST) to measure the ion temperature and fast electron fluxes. A Langmuir probe assembly was added on the top of the RFA head to control the RFA position relative to the last closed flux surface and to have a possibility to measure the electron density and temperature as well. Except the ion temperature, the fast electron fluxes from both ion and electron drift sides have been measured during lower hybrid current drive. The RFA probe has been also used to measure the fast electrons associated with edge localized modes (ELMs), indicating their substantial presence in the scrape-off-layer plasma of EAST.

Interactive WebGL-based 3D visualizations for EAST experiment

In recent years EAST (Experimental Advanced Superconducting Tokamak) experimental data are being shared and analyzed by an increasing number of international collaborators. Developing a user-friendly interface to visualize the data, meta data and the relevant parts of the device is becoming more and more important to aid scientists and engineers. Compared with the previous virtual EAST system based on VRML/Java3D [1] (Li et al., 2014), a new technology is being adopted to create a 3D visualization system based on HTML5 and WebGL, which runs without the need for plug-ins or third party components. The interactive WebGL-based 3D visualization system is a web-portal integrating EAST 3D models, experimental data and plasma videos. It offers a highly interactive interface allowing scientists to roam inside EAST device and view the complex 3-D structure of the machine. It includes technical details of the device and various diagnostic components, and provides visualization of diagnostic metadata with a direct link to each signal name and its stored data. In order for the quick access to the device 3D model, the original CAD model was discretized into different layers with different simplification. It allows users to search for plasma videos in any experiment and analyze the video frame by frame. In this paper, we present the implementation details to enable realistic rendering and improve performance.

The Accomplishments and Next-Step Plan of EAST in Support of Fusion

The experimental advanced superconducting tokamak (EAST) is a superconducting tokamak, which successfully achieved the first plasma discharge in 2006. The major radius of EAST plasma is 1.9 m, and its central magnet field is 3.5 T. In the past few years, EAST has made many achievements, such as 60-s double-null divertor configuration plasma and 1-MA plasma. In 2012, 411-s long-pulse discharge with 0.28-MA plasma current and 32-s H-mode operation was achieved at the seventh campaign. In addition, various means for mitigating ELMs have also been demonstrated to facilitate long-pulse operation. The experimental results and engineering experience obtained from EAST can be used as a reference for fusion next step. To make further progress, much optimization work should be done in next step. Based on the present EAST heating system, two sets of 4-MW neutral beam injection, one set of electron cyclotron resonance heating, and lower hybrid current drive will be installed to achieve more than 30-MW total heating power. The upper divertor will be updated to W–Cu divertor, which consists of monoblock structure and tungsten armor that can withstand 10- <formula formulatype="inline" xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"><tex Notation="TeX">${\rm MW}\cdot{\rm m}^{-2}$</tex></formula> heat load. To increase the upper divertor's ash removal efficiency, one new cryopump will be added on the behind of it. Aiming to improve plasma configuration and enhance its stabilization, the EAST resonant magnetic perturbation (RMP) coils will be designed, fabricated, and installed. The RMP coils will integrate the functions of error field correction, edge localized mode, and resistive wall mode. About 45 kinds of diagnostics will be used in the next campaign and all of them will be integrated on six ports as port plug. In addition, some fusion technologies are also considered in EAST to validate for ITER and future reactor.

The implementation of magnetics verification in EAST plasma control system

Experimental advanced superconducting tokamak (EAST) is an experimental device aiming at steady state plasma operation for fusion research. The values of many discharge parameters, such as plasma shape, position and current must be directly acquired or indirectly evaluated from the magnetic measurements, so the accuracy of magnetic measurements plays an important role in reliable plasma control performance. A method for verifying the key magnetic measurements in real time for each shot is described in this paper. Such magnetics verification will prevent the discharge from a key magnetic signal failure and ensure the quality of a successful discharge. The diagnostics verification algorithm has been implemented in the plasma control system for the EAST. The implementation details and its application in the recent experiment are presented in this paper.

Neutral beam injection simulation of EAST

EAST (Experimental Advanced Superconducting Tokamak) is a fully superconducting tokamak with single/double null divertor configuration. The NBI (Neutral Beam Inject) system of EAST has two deuterium beam lines, in which beam power is 4 MW and energy is 80 keV. Neutral beam heating in EAST tokamak is simulated using NUBEAM code. The NUBEAM module is a Monte Carlo modeling of fast ion species in an axisymmetric tokamak. With the numerical simulation, the scenarios of the future neutral beam heating in EAST plasma are predicted. The influences of the injection angle on the neutral beam heating, current drive and shine-through power are analyzed by the simulation.

EAST plasma control system

Experimental advanced superconducting tokamak (EAST), the first tokomak with ITER-like fully superconductive poloidal and toroidal coils, successfully completed its engineering commission in March 2006; had its first plasma in September 2006 [B.N. Wan, et al., The first plasma of EAST, in: 21st IAEA Fusion Energy Conference, Chengdu, China, 16–21 October, 2007. [1]] and further generated diverted plasma in January 2007. The plasma control system was adapted from DIII-D plasma control system architecture and jointly developed by the DIII-D and EAST plasma control team. After briefly outlining the hardware and software architecture of the EAST plasma control system, this paper summarizes the plasma control results in the first plasma and first diverted plasma campaigns.

Overview of diagnostic system in the experimental advanced superconducting tokamak

The first plasma was achieved in the experimental advanced superconducting tokamak (EAST) on September 26, 2006. Single-and double-null diverter plasma discharges were also achieved successfully in the EAST in January 2007. The plasma diagnostics for the first EAST plasma study have been developed as follows: a vertical one-channel far-infrared hydrogen cyanide (HCN) laser interferometer for measuring the line average density, a ten-channel soft X-ray array for intensity measurement, a 16-channel heterodyne electric cyclotron emission for measuring the profile of the electron temperature, an eight channel XUV bolometer array to measure plasma radiation losses, a three-channel hard X-ray array for intensity measurement, an electromagnetic measurement system, a 35-channel Hα radiation array, 20 triple probes for diverter plasma, a one-channel visible bremsstrahlung emission, an impurity optical spectrum measurement system, and two optical spectroscopic multichannel analyzers. The first experimental results of the diagnostic system and the plan of what is to come are summarized in this paper.

Thermal and mechanical analysis of the EAST plasma facing components

The Experimental Advanced Superconducting Tokamak (EAST) is an advanced steady-state plasma physics experimental device, which has a long pulse (60–1000 s) capability, a flexible poloidal field system, auxiliary heating and current drive systems. Consequently, it will be able to accommodate divertor heat loads that make it an attractive test for the development of advanced tokamak operating modes. Now, the engineering designs for the EAST plasma facing component (PFC) are in progress. All the PFCs are made of copper alloy (CuCrZr) on which graphite tiles are mechanically attached. A layer of flexible graphite is placed between graphite tiles and copper alloy plate to ensure the better thermal transfer. The copper alloy are actively cooled by water flowing the cooling channel drilled by a special technology. This paper will introduce the thermal analysis including the mechanical behavior analysis for the plasma facing component using finite element method. Some candidate joint structure among the graphite tile, flexible graphite sheet and copper alloy are discussed in this paper.