Predictive Simulations of Plasma Heating and Current Drive by Neutral Beam Injection on EAST

Long pulse and high performance steady-state operation is the main scientific mission of experimental advanced superconducting tokamak (EAST). In order to achieve this objective, high-power auxiliary heating systems are essential. Radio frequency (RF) wave heating and neutral beam injection (NBI) are two principal methods. NBI is an effective method of plasma heating and current drive, and it has been used in many magnetic confinement fusion devices. Based on the plasma equilibrium of EAST (Li et al., Plasma Phys Control Fusion 55:125008, 2013) plus previous EAST experimental data used as initial conditions, the NBI module (Polevoi et al., JAERI-Data, 1997) employed in automated system for transport analysis (ASTRA) code (Pereverzev et al., IPP-Report, 2002) is applied to predict the effects of plasma heating and current drive with different neutral beam injection power levels. At certain levels of plasma densities and plasma current densities, the simulation results show that the NBI heats plasma effectively, also increases the proportions of NB current and bootstrap current among total current significantly.     

First experimental results of intrinsic torque on EAST

Direct measurements of the intrinsic torque profile in L-mode plasmas on the EAST tokamak have been performed using the balanced neutral beam injection. Co-and counter-current neutral beams are modulated to balance the intrinsic and externally injected torques, which result in the rotation profile close to zero and flat. The experimental results show that the intrinsic torque derived from momentum balance equations is found to be in the co-current direction, peaked in the plasma edge and negligibly small in the core.     

Numerical Simulation of Non-Inductive Current Driven Scenario in EAST Using Neutral Beam Injection

For achieving the scientific mission of long pulse and high performance operation,experimental advanced superconducting tokamak(EAST) applies fully superconducting magnet technology and is equiped with high power auxiliary heating system.Besides RF(Radio Frequency) wave heating,neutral beam injection(NBI) is an effective heating and current drive method in fusion research.NBCD(Neutral Beam Current Drive) as a viable non-inductive current drive source plays an important role in quasi-steady state operating scenario for tokamak.The non-inductive current driven scenario in EAST only by NBI is predicted using the TSC/NUBEAM code.At the condition of low plasma current and moderate plasma density,neutral beam injection heats the plasma effectively and NBCD plus bootstrap current accounts for a large proportion among the total plasma current for the flattop time.     

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.     

EAST托卡马克的中性束注入方案

高能中性束注入(Neutral beam injection,NBI)是核聚变装置托卡马克采用的芯部辅助加热和非感应电流驱动主要手段之一.本文介绍了国家大科学工程全超导托卡马克实验装置(Experimental advanced super-conductingtokamak,EAST)上的高能NBI加热方案及注入器的工程要求,并讨论了中性束在EAST等离子体中的传输等相关问题.     

Thermal analysis of EAST neutral beam injectors for long-pulse beam operation

Two sets of neutral beam injectors(NBI-1 and NBI-2) have been mounted on the EAST tokamak since 2014. NBI-1 and NBI-2 are co-direction and counter-direction, respectively. As with indepth physics and engineering study of EAST, the ability of long pulse beam injection should be required in the NBI system. For NBIs, the most important and difficult thing that should be overcome is heat removal capacity of heat loaded components for long-pulse beam extraction. In this article, the thermal state of the components of EAST NBI is investigated using water flow calorimetry and thermocouple temperatures. Results show that(1) operation parameters have an obvious influence on the heat deposited on the inner components of the beamline,(2) a suitable operation parameter can decrease the heat loading effectively and obtain longer beam pulse length, and(3) under the cooling water pressure of 0.25 MPa, the predicted maximum beam pulse length will be up to 260 s with 50 keV beam energy by a duty factor of 0.5. The results present that, in this regard, the EAST NBI-1 system has the ability of long-pulse beam injection.     

Development of a Data Acquisition Control System for the First NBI on EAST

Neutral beam injection （NBI） heating is one of the most efficient auxiliary plasma heating methods for fusion devices. The data acquisition control system （DACS） with PXI （pe- ripheral component interconnect extensions for instrumentation） data acquisition card for the first NBI system in the experimental advanced superconducting tokamak （EAST） is presented in this paper. As an important sub-system, DACS is designed to obtain physical measurement signals in the EAST NBI system and to deal and store these data with the Lempel-Ziv-Oberhumer （LZO） lossless data compression algorithm, as well as offer convenient data call-back and access inter- faces to the user for examining and analyzing the data. Experimental results show that accurate data will ensure that researchers correctly analyze it and then properly adjust the experimental parameters or operation, so DACS should take a large step in improving experimental efficiency. Tile hardware and software sections are briefly presented in this paper, and now this system has been tested to be able to work reliably and steadily.     

Analysis of Power Deposition on Heat Load Components for EAST Neutral Beam Injector

Neutral beam injection is recognized as one of the most effective means of plasma heating. The target values of EAST Neutral beam injector (NBI) are beam energy 50–80 keV, injection beam total power 2–4 MW, beam pulse width 10–100 s. The beam power will deposit on the beam collimator due to the beam divergence and it will cause heat damage to heat load components, or even destroy the entire NBI system. In order to decrease the risk, the beam power deposited on heat load components should be assessed. In this article, the percent of power deposition on each heat load components has been calculated using Gaussian beam transmission model. Comparison of the results measured with water flow calorimeter and calculated results shows the beam transmission model has relative good agreement with real distribution. The results can direct the operation parameter optimization of EAST NBI.     

Simulations of first-orbit losses of neutral beam injection(NBI) fast ions on EAST

Simulations of first-orbit losses of neutral beam injection(NBI) fast ions in the EAST tokamak have been studied in detail by using the orbit-following code GYCAVA and the NBI code TGCO. Beam ion losses with the wall boundary are smaller than those with the last closed flux surface boundary. In contrast to heat loads on the wall without radio frequency wave(RFW)antennas, heat loads on the wall with RFW antennas are distributed more locally near the RFW antennas. The direction of the toroidal magnetic field dramatically affects the final positions of lost fast ions, which is related to the magnetic drift. The numerical results on heat loads of beam ions corresponding to different toroidal magnetic fields are qualitatively consistent with the experimental results. Beam ion losses increase with the beam energy for the co-current NBIs and the counter-perpendicular NBI. We have studied the behavior of fast ions produced by a small section neutral beam(beamlet) by using the numerical tool NBIT. The distributions of the loss fraction of beamlet fast ions peaked near the edge of the beam section for the counter-current NBIs, and they are related to the injection angle. This indicates that the first-orbit losses can be reduced by changing the shape of beam cross section.     

Data Acquisition System of Neutral Beam Injector on EAST

In order to supervise the elements of the neutral beam injector (NBI) spatially located at several places, a distributed NBI data acquisition system (NBIDAS) on experimental advanced superconducting tokamak (EAST) is developed in this paper. NBIDAS consists of field instrument and measurement devices, servers and remote data processing terminals. In order to remotely manage and monitor the field devices of the NBI system, a device management client software is also developed as the human–machine interfaces between the field devices and remote system administrators. A control signal acquisition system is developed for diagnosing these generated analog and digital signals from the NBI control system. NBIDAS based on network technologies is capable of extending system functions and upgrading devices. The detail of the architecture and implementation of the NBIDAS on EAST is discussed in the paper.     

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.     

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.     

全超导非圆截面托卡马克中性束注入窗口材料活化研究

中性束注入加热为全超导非圆截面托卡马克(EAST)主要辅助加热方式之一.伴随着中性束注入加热,等离子体中子出射强度可达到1014 n/s.由于中性束注入窗口具有较大的开口尺寸,窗口泄露的大量中子可能影响系统的安全稳定运行.本文基于EAST中性束二维模型和蒙特卡洛程序MCNP与材料活化程序FISPACT,研究EAST两条...     

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.     

The Development of Data Acquisition and Remote Real-Time Display System for EAST NBI

The data acquisition and remote real-time display system for the neutral beam injectors (NBI) on experimental advanced superconducting tokamak (EAST) are described in this paper. Distributed computer systems including local data acquisition (DAQ) facility, remote data server (DS), real-time display terminal are adopted with Linux and Windows operating system. Experimental signals are gathered by DAQ device at local working field. On the one hand, these gathered data will be sent to DS which runs on remote server main control layer on EAST NBI control network for saving and processing; on the other hand, these data will be sent to real-time display terminal which runs on remote monitoring layer on EAST NBI for displaying and monitoring experimental signals real-timely. Another point needs to be mentioned is that the real-time display software can call back historical data from DS for querying. The software of data acquisition and DS are programmed by C language while the real-time display software is programmed by Labview flow chart. The hardware mainly includes DAQ cards, server, industrial personal computer and others auxiliary hardware. Now the system proved to be performed well through experiments on NBI testing bed.     

The Influence of Neutral Beam Injection on the Heating and Current Drive with Electron Cyclotron Wave on EAST

Both neutral beam injection(NBI) and electron cyclotron resonance heating(ECRH) have been applied on the Experimental Advanced Superconducting Tokamak(EAST)in the 2015 campaign. In order to achieve more effective heating and current drive, the effects of NBI on the heating and current drive with electron cyclotron wave(ECW) are analyzed utilizing the code TORAY and experimental data in the shot #54411 and #54417. According to the experimental and simulated results, for the heating with ECW, NBI can improve the heating efficiency and move the power deposition place towards the inside of the plasma. On the other hand, for the electron cyclotron current drive(ECCD), NBI can also improve the efficiency of ECCD and move the place of ECCD inward. These results will be valuable for the center heating, the achievement of fully non-inductive current drive operation and the suppression of magnetohydrodynamic(MHD) instabilities with ECW on EAST or ITER with many auxiliary heating methods.     

Water Flow Calorimetry System of EAST Neutral Beam Injector

Neutral beam injection (NBI) is recognized as one of the most effective means of plasma heating. The EAST NBI water flow calorimetry system (WFCS) based on PCI extensions for instrumentation (PXI) was established, it can measure temperature rise and flow rate of cooling water of the heat load components, and achieve beam power distribution and neutralization efficiency. Experimental data obtained from WFCS are feedback of the ion source operation state and direct the operation parameter optimization of the ion source. Experimental results show that the WFCS is stable, reliable, and meet the experimental requirements fully.     

Analysis of Power Distribution on Beamline Components at Different Neutralization Efficiencies on NBI Test Stand

Neutral beam injection is recognized as one of the most effective means for plasma heating. According to the research plan of the EAST physics experiment, two sets of neutral beam injector(4–8 MW, 10–100 s) were built and operated in 2014. Neutralization efficiency is one of the important parameters for neutral beam. High neutralization efficiency can not only improve injection power at the same beam energy, but also decrease the power deposited on the heat-load components in the neutral beam injector(NBI). This research explores the power deposition distribution at different neutralization efficiencies on the beamline components of the NBI device. This work has great significance for guiding the operation of EAST-NBI, especially in long pulse and high power operation, which can reduce the risk of thermal damage of the beamline components and extend the working life of the NBI device.     

Ignition Approach by Neutral Beam Injection Heating in Impurity Contaminated Tokamak Reactors

Available heating power by neutral beam injection in a tokamak reactor is evaluated semi-empirically. Using this estimated value, device and plasma parameters to ignite the plasma in impurity contaminated tokamak reactors are investigated. By lowering the plasma density and concurrently by enlarging the plasma minor radius or aspect ratio, the difficulty of NBI heating can be avoided, and the ignition is almost always possible both for trapped ion mode and Alcator scaling laws.     

Numerical prediction of pellet injection effects on core plasma fueling in the KSTAR tokamak

Since pellet injection into tokamak plasmas has been found to be an effective method for fueling and profile modification of core plasmas in tokamak experiments, a hypothetical injection of deutrium pellets into the KSTAR tokamak is numerically simulated in this work to investigate its influences on the fueling and transport of the core plasma depending on pellet parameters. A neutral gas shielding model and a pellet drift displacement model are used to describe the ablation and mass deposition from pellets on core plasma profiles. These models are coupled with a 1.5-dimensional (1.5D) core transport code to calculate the plasma density and temperature profiles responding to pellets injected into the target plasma. The simulation results indicate that a HFS (high field side) injection achieves more effective fueling due to a deeper pellet penetration into the core plasma, compared with a LFS (low field side) injection. The plasma density is found to increase during sequential pellet injections from both HFS and LFS, but the HFS case shows better fueling performance owing to a drift of the pellet ablatant in the major radius direction resulting in the deeper pellet penetration. Increasing the size and injection velocity of the pellet contributes to enhance the fueling efficiency. However, raising the power of neutral beam injection heating reduces the fueling efficiency because the pellet mass deposition is shifted toward the edge region in high temperature plasmas. It is concluded that the pellet size and injection direction among pellet and plasma parameters have the most dominant effects on fueling performance while the pellet velocity and heating power have relatively small influences on fueling.