Toroidal torques,generated by the resonant magnetic perturbation(RMP)and acting on the plasma column,are numerically systematically investigated for an ITER baseline scenario.The neoclassical toroidal viscosity(NTV),i...Toroidal torques,generated by the resonant magnetic perturbation(RMP)and acting on the plasma column,are numerically systematically investigated for an ITER baseline scenario.The neoclassical toroidal viscosity(NTV),in particular the resonant portion,is found to provide the dominant contribution to the total toroidal torque under the slow plasma flow regime in ITER.While the electromagnetic torque always opposes the plasma flow,the toroidal torque associated with the Reynolds stress enhances the plasma flow independent of the flow direction.A peculiar double-peak structure for the net NTV torque is robustly computed for ITER,as the toroidal rotation frequency is scanned near the zero value.This structure is found to be ultimately due to a non-monotonic behavior of the wave-particle resonance integral(over the particle pitch angle)in the superbanana plateau NTV regime in ITER.These findings are qualitatively insensitive to variations of a range of factors including the wall resistivity,the plasma pedestal flow and the assumed frequency of the rotating RMP field.展开更多
This study investigates the influence of runaway current in runaway plasmas on the dynamics of sawtooth oscillations and resultant loss of runaway electrons(RE)using the 3D magnetohydrodynamic(MHD)code M3D-C^(1)(Jardi...This study investigates the influence of runaway current in runaway plasmas on the dynamics of sawtooth oscillations and resultant loss of runaway electrons(RE)using the 3D magnetohydrodynamic(MHD)code M3D-C^(1)(Jardin et al 2012 J.Comput.Sci.Discovery 6014002).Using an HL-2A-like equilibrium,we confirm that in the linear phase,the impact of REs on resistive internal kink instabilities is consistent with previous research.In the nonlinear phase,as the runaway current fully replaces the plasmas current,we observe a significant suppression of sawtooth oscillations,with the first sawtooth cycle occurring earlier compared to the case without runaway current.Following the first sawtooth collapse,plasma current density,runaway current density,and safety factor(q)flatten within the q=1 surface,albeit displaying fine structures.Subsequently,the growing high torodial(n)and poloidal(m)mode number modes disrupt the magnetic surfaces,leading to the loss of REs outside the q=1 surface,while minimally affecting the majority of REs well-confined within it.Thus,in the current model,the physical processes associated with the presence of sawtooth oscillations do not effectively dissipate runaway current,as REs are assumed to be collisionless.In addition,the final profile of runaway current density exhibits increased steepening near the q=1 surface in contrast to the initial profile,displaying a distinctive corrugated inhomogeneity influenced by the growing fluctuation of the n=0 component.Finally,detailed convergence tests are conducted to validate the numerical simulations.展开更多
Effects of three-dimensional(3D)magnetic field perturbations due to feedback control of an unstable n=1(n is toroidal mode number)resistive wall mode(RWM)on the energetic particle(EP)losses are systematically investig...Effects of three-dimensional(3D)magnetic field perturbations due to feedback control of an unstable n=1(n is toroidal mode number)resistive wall mode(RWM)on the energetic particle(EP)losses are systematically investigated for the HL-3 tokamak.The MARS-F(Liu et al 2000 Phys.Plasmas 73681)code,facilitated by the test particle guiding center tracing module REORBIT,is utilized for the study.The RWM is found to generally produce no EP loss for cocurrent particles in HL-3.Assuming the same perturbation level at the sensor location for the close-loop system,feedback produces nearly the same loss of counter-current EPs compared to the open-loop case.Assuming however that the sensor signal is ten times smaller in the close-loop system than the open-loop counter part(reflecting the fact that the RWM is more stable with feedback),the counter-current EP loss is found significantly reduced in the former.Most of EP losses occur only for particles launched close to the plasma edge,while particles launched further away from the plasma boundary experience much less loss.The strike points of lost EPs on the HL-3 limiting surface become more scattered for particles launched closer to the plasma boundary.Taking into account the full gyro-orbit of particles while approaching the limiting surface,REORBIT finds slightly enhanced loss fraction.展开更多
Many magnetohydrodynamic stability analyses require generation of a set of equilibria with a fixed safety factor q-profile while varying other plasma parameters.A neural network(NN)-based approach is investigated that...Many magnetohydrodynamic stability analyses require generation of a set of equilibria with a fixed safety factor q-profile while varying other plasma parameters.A neural network(NN)-based approach is investigated that facilitates such a process.Both multilayer perceptron(MLP)-based NN and convolutional neural network(CNN)models are trained to map the q-profile to the plasma current density J-profile,and vice versa,while satisfying the Grad–Shafranov radial force balance constraint.When the initial target models are trained,using a database of semianalytically constructed numerical equilibria,an initial CNN with one convolutional layer is found to perform better than an initial MLP model.In particular,a trained initial CNN model can also predict the q-or J-profile for experimental tokamak equilibria.The performance of both initial target models is further improved by fine-tuning the training database,i.e.by adding realistic experimental equilibria with Gaussian noise.The fine-tuned target models,referred to as fine-tuned MLP and fine-tuned CNN,well reproduce the target q-or J-profile across multiple tokamak devices.As an important application,these NN-based equilibrium profile convertors can be utilized to provide a good initial guess for iterative equilibrium solvers,where the desired input quantity is the safety factor instead of the plasma current density.展开更多
In this paper, gas control on EAST in open and closed loop is discussed and its implementation into EASTPCS （plasma control system for the experimental advanced supercon- ducting tokamak） is introduced. Using a mode...In this paper, gas control on EAST in open and closed loop is discussed and its implementation into EASTPCS （plasma control system for the experimental advanced supercon- ducting tokamak） is introduced. Using a model to describe the plasma density response to the gas puff command, a gas simulation server （simserver） using MATLAB simulink tools and real time workshop was built up. Proper operation of the gas control algorithm was verified using this simserver. The simulation results suggested that the gas control can be applied in the next EAST campaign.展开更多
The EAST plasma control system （PCS） is in continuous development to satisfy the EAST experimental requirements. In order to realize low latency and distortion-free signal transmission between PCS and servo systems ...The EAST plasma control system （PCS） is in continuous development to satisfy the EAST experimental requirements. In order to realize low latency and distortion-free signal transmission between PCS and servo systems such as the poloidal field power supply, in-vessel coil power supply and real-time scope, reflective memory boards （RFM） were applied. The new hardware layout and enhanced performance are reported. Newly implemented PCS control algorithms for gas control and real-time data display are also presented.展开更多
Reconstruction of experimental axisymmetric equilibria is an important part of tokamak data analysis. Fourier expansion is applied to reconstruct the vessel current distribution in EFIT code. Benchmarking and testing ...Reconstruction of experimental axisymmetric equilibria is an important part of tokamak data analysis. Fourier expansion is applied to reconstruct the vessel current distribution in EFIT code. Benchmarking and testing calculations are performed to evaluate and validate this algorithm. Two cases for circular and non-circular plasma discharges are presented. Fourier expansion used to fit the eddy current is a robust method and the real time EFIT can be introduced to the plasma control system in the coming campaign.展开更多
A long pulse electron cyclotron resonance heating（ECRH）system has been developed to meet the requirements of steady-state operation for the EAST superconducting tokamak,and the first EC wave was successfully injecte...A long pulse electron cyclotron resonance heating（ECRH）system has been developed to meet the requirements of steady-state operation for the EAST superconducting tokamak,and the first EC wave was successfully injected into plasma during the 2015 spring campaign.The system is mainly composed of four 140 GHz gyrotron systems,4 ITER-Like transmission lines,4 independent channel launchers and corresponding power supplies,a water cooling,control ＆inter-lock system etc.Each gyrotron is expected to deliver a maximum power of 1 MW and be operated at 100-1000 s pulse lengths.The No.1 and No.2 gyrotron systems have been installed.In the initial commissioning,a series of parameters of 1 MW 1 s,900 k W 10 s,800 k W 95 s and650 k W 753 s have been demonstrated successfully on the No.1 gyrotron system based on calorimetric dummy load measurements.Significant plasma heating and MHD instability suppression effects were observed in EAST experiments.In addition,high confinement（H-mode）discharges triggered by ECRH were obtained.展开更多
In the 2016 EAST experimental campaign,a steady-state long-pulse H-mode discharge with an ITER-like tungsten divertor lasting longer than one minute has been obtained using only RF heating and current drive,through an...In the 2016 EAST experimental campaign,a steady-state long-pulse H-mode discharge with an ITER-like tungsten divertor lasting longer than one minute has been obtained using only RF heating and current drive,through an integrated control of the wall conditioning,plasma configuration,divertor heat flux,particle exhaust,impurity management,and effective coupling of multiple RF heating and current drive sources at high injected power.The plasma current（Ip - 0.45 MA） was fully-noninductively driven（Vloop 〈 0.0 V） by a combination of-2.5 MW LHW,-0.4 MW ECH and -0.8 MW ICRF.This result demonstrates the progress of physics and technology studies on EAST,and will benefit the physics basis for steady state operation of ITER and CFETR.展开更多
In 2021,EAST realized a steady-state long pulse with a duration over 100 s and a core electron temperature over 10 keV.This is an integrated operation that resolves several key issues,including active control of wall ...In 2021,EAST realized a steady-state long pulse with a duration over 100 s and a core electron temperature over 10 keV.This is an integrated operation that resolves several key issues,including active control of wall conditioning,long-lasting fully noninductive current and divertor heat/particle flux.The fully noninductive current is driven by pure radio frequency(RF)waves with a lower hybrid current drive power of 2.5 MW and electron cyclotron resonance heating of 1.4 MW.This is an excellent experimental platform on the timescale of hundreds of seconds for studying multiscale instabilities,electron-dominant transport and particle recycling(plasma-wall interactions)under weak collisionality.展开更多
A new method, by using eigenmodes to reduce the fitting parameters and precalculated eddy current based on a lump parameter circuit equation, is applied to reconstruct the vacuum field for EAST plasma startup.
Magnetohydrodynamic （MHD） n=1 kink mode with n the toroidal mode number is studied and the operational beta limit, constrained by the mode, is calculated for the equilibrium of HL-2A by using the GATO code. Approxim...Magnetohydrodynamic （MHD） n=1 kink mode with n the toroidal mode number is studied and the operational beta limit, constrained by the mode, is calculated for the equilibrium of HL-2A by using the GATO code. Approximately the same beta limit is obtained for configurations with a value of the axial safety factor q0 both larger and less than 1. Without the stabilization of the conducting wall, the beta limit is found to be 0.821% corresponding to a normalized beta value of βN^c=2.56 for a typical HL-2A discharge with a plasma current Ip=0.245 MA, and the scaling of βN^c -constant is confirmed.展开更多
The experimental advanced superconducting tokamak （EAST） is the first full superconducting tokamak with a D-shaped cross-sectional plasma presently in operation. Its poloidal coils are relatively far from the plasma...The experimental advanced superconducting tokamak （EAST） is the first full superconducting tokamak with a D-shaped cross-sectional plasma presently in operation. Its poloidal coils are relatively far from the plasma due to the necessary thermal isolation from the superconducting magnets, which leads to relatively weaker coupling between plasma and poloidal field. This may cause more difficulties in controlling the vertical instability by using the poloidal coils. The measured growth rates of vertical stability are compared with theoretical calculations, based on a rigid plasma model. Poloidal beta and internal inductance are varied to investigate their effects on the stability margin by changing the values of parameters αn and γn（Howl et al 1992 Phys. Fluids B 4 1724）, with plasma shape fixed to be a configuration with k = 1.9 and 5 = 0.5. A number of ways of studying the stability margin are investigated. Among them, changing the values of parameters κ and li is shown to be the most effective way to increase the stability margin. Finally, a guideline of stability margin Ms（ki, li, A） to a new discharge scenario showing whether plasmas can be stabilized is also presented in this paper.展开更多
Lower hybrid simulation code （LSC） is integrated into transport code ONETWO to model discharges of EAST with heating and current drive using both lower hybrid wave and neutral beam injection. To test the integration...Lower hybrid simulation code （LSC） is integrated into transport code ONETWO to model discharges of EAST with heating and current drive using both lower hybrid wave and neutral beam injection. To test the integration, the current profiles driven by both lower hybrid wave and neutral beam injection are presented for a given equilibrium and sample density and temperature profiles before carrying out a time advancing simulation. For a steady state H-mode discharge, the temperature profiles are evolved using GLF23 transport model with a boundary value specified at the normalized minor radius pN = 0.9. A case of steady state with non-inductively driven current accounting for up to 80% is presented.展开更多
Blast-wave-driven hydrodynamic instabilities are studied in the presence of a background B-field through experiments and simulations in the high-energy-density(HED)physics regime.In experiments conducted at the Labora...Blast-wave-driven hydrodynamic instabilities are studied in the presence of a background B-field through experiments and simulations in the high-energy-density(HED)physics regime.In experiments conducted at the Laboratoire pour l’utilisation des lasers intenses(LULI),a laserdriven shock-tube platform was used to generate a hydrodynamically unstable interface with a prescribed sinusoidal surface perturbation,and short-pulse x-ray radiography was used to characterize the instability growth with and without a 10-T B-field.The LULI experiments were modeled in FLASH using resistive and ideal magnetohydrodynamics(MHD),and comparing the experiments and simulations suggests that the Spitzer model implemented in FLASH is necessary and sufficient for modeling these planar systems.These results suggest insufficient amplification of the seed B-field,due to resistive diffusion,to alter the hydrodynamic behavior.Although the ideal-MHD simulations did not represent the experiments accurately,they suggest that similar HED systems with dynamic plasma-β(=2μ_(0)ρv^(2)/B^(2))values of less than∼100 can reduce the growth of blast-wave-driven Rayleigh–Taylor instabilities.These findings validate the resistive-MHD FLASH modeling that is being used to design future experiments for studying B-field effects in HED plasmas.展开更多
Targets for low-adiabat direct-drive-implosion experiments on OMEGA must meet rigorous specifications and tight tolerances on the diameter,wall thickness,wall-thickness uniformity,and presence of surface features.Of t...Targets for low-adiabat direct-drive-implosion experiments on OMEGA must meet rigorous specifications and tight tolerances on the diameter,wall thickness,wall-thickness uniformity,and presence of surface features.Of these,restrictions on the size and number of defects(bumps and depressions)on the surface are the most challenging.The properties of targets that are made using vapor-deposition and solution-based microencapsulation techniques are reviewed.Targets were characterized using confocal microscopy,bright-and dark-field microscopy,atomic force microscopy,electron microscopy,and interferometry.Each technique has merits and limitations,and a combination of these techniques is necessary to adequately characterize a target.The main limitation with the glow-discharge polymerization(GDP)method for making targets is that it produces hundreds of domes with a lateral dimension of 0.7-2 μm.Polishing these targets reduces the size of some but not all domes,but it adds scratches and grooves to the surface.Solution-made polystyrene shells lack the dome features of GDP targets but have hundreds of submicrometer-size voids throughout the wall of the target;a few of these voids can be as large as~12 μm at the surface.展开更多
For the advanced tokamak,the particle deposition and thermal load on the divertor is a big challenge.By moving the strike points on divertor target plates,the position of particle deposition and thermal load can be sh...For the advanced tokamak,the particle deposition and thermal load on the divertor is a big challenge.By moving the strike points on divertor target plates,the position of particle deposition and thermal load can be shifted.We could adjust the Poloidal Field（PF） coil current to achieve the strike point position feedback control.Using isoflux control method,the strike point position can be controlled by controlling the X point position.On the basis of experimental data,we establish relational expressions between X point position and strike point position.Benchmark experiments are carried out to validate the correctness and robustness of the control methods.The strike point position is successfully controlled following our command in the EAST operation.展开更多
Transport of fast ions is a crucial issue during the operation of ITER.Redistribution of neutral beam injection(NBI)fast ions by the ideal internal magnetohydrodynamic(MHD)instabilities in ITER is studied utilizing th...Transport of fast ions is a crucial issue during the operation of ITER.Redistribution of neutral beam injection(NBI)fast ions by the ideal internal magnetohydrodynamic(MHD)instabilities in ITER is studied utilizing the guiding-center code ORBIT(White R B and Chance M S 1984Phys.Fluids 272455).Effects of the perturbation amplitude A of the internal kink,the perturbation frequency f of the fishbone instability,and the toroidal mode number n of the internal kink are investigated,respectively,in this work.The n=1 internal kink mode can cause NBI fast ions transporting in real space from regions of 0<s≤0.32 to 0.32<s≤0.53,where s labels the normalized plasma radial coordinate.The transport of fast ions is greater as the perturbation amplitude increases.The maximum relative change of the number of fast ions approaches 5%when the perturbation amplitude rises to 500 G.A strong transport is generated between the regions of 0<s≤0.05 and 0.05<s≤0.12 in the presence of the fishbone instability.Higher frequency results in greater transport,and the number of fast ions in 0<s≤0.05 is reduced by 30%at the fishbone frequency of 100 k Hz.Perturbations with higher n will lead to the excursion of fast ion transport regions outward along the radial direction.The loss of fast ions,however,is not affected by the internal MHD perturbation.Strong transport from 0<s≤0.05 to 0.05<s≤0.12 does not influence the plasma heating power of ITER,since the NBI fast ions are still located in the plasma core.On the other hand,the influence of fast ion transport from 0<s≤0.32 to 0.32<s≤0.53 needs further study.展开更多
基金funded by National Natural Science Foundation of China(NSFC)(Nos.12075053,11505021 and 11975068)by National Key R&D Program of China(No.2022YFE 03060002)+1 种基金by Fundamental Research Funds for the Central Universities(No.2232024G-10)supported by the U.S.DoE Office of Science(No.DE-FG02–95ER54309)。
文摘Toroidal torques,generated by the resonant magnetic perturbation(RMP)and acting on the plasma column,are numerically systematically investigated for an ITER baseline scenario.The neoclassical toroidal viscosity(NTV),in particular the resonant portion,is found to provide the dominant contribution to the total toroidal torque under the slow plasma flow regime in ITER.While the electromagnetic torque always opposes the plasma flow,the toroidal torque associated with the Reynolds stress enhances the plasma flow independent of the flow direction.A peculiar double-peak structure for the net NTV torque is robustly computed for ITER,as the toroidal rotation frequency is scanned near the zero value.This structure is found to be ultimately due to a non-monotonic behavior of the wave-particle resonance integral(over the particle pitch angle)in the superbanana plateau NTV regime in ITER.These findings are qualitatively insensitive to variations of a range of factors including the wall resistivity,the plasma pedestal flow and the assumed frequency of the rotating RMP field.
基金supported in part by the National Key R&D Program of China (No.2022YFE03040002)the Natural Science Foundation of Sichuan (No.2022NSFSC1814)+3 种基金National Natural Science Foundation of China (Nos.12305246,12175053 and 12261131622)the Italian Ministry of Foreign Affairs (No.CN23GR02)the Fundamental Research Funds for the Central Universitiessupported by US Department of Energy (No.DE-AC0209CH11466)。
文摘This study investigates the influence of runaway current in runaway plasmas on the dynamics of sawtooth oscillations and resultant loss of runaway electrons(RE)using the 3D magnetohydrodynamic(MHD)code M3D-C^(1)(Jardin et al 2012 J.Comput.Sci.Discovery 6014002).Using an HL-2A-like equilibrium,we confirm that in the linear phase,the impact of REs on resistive internal kink instabilities is consistent with previous research.In the nonlinear phase,as the runaway current fully replaces the plasmas current,we observe a significant suppression of sawtooth oscillations,with the first sawtooth cycle occurring earlier compared to the case without runaway current.Following the first sawtooth collapse,plasma current density,runaway current density,and safety factor(q)flatten within the q=1 surface,albeit displaying fine structures.Subsequently,the growing high torodial(n)and poloidal(m)mode number modes disrupt the magnetic surfaces,leading to the loss of REs outside the q=1 surface,while minimally affecting the majority of REs well-confined within it.Thus,in the current model,the physical processes associated with the presence of sawtooth oscillations do not effectively dissipate runaway current,as REs are assumed to be collisionless.In addition,the final profile of runaway current density exhibits increased steepening near the q=1 surface in contrast to the initial profile,displaying a distinctive corrugated inhomogeneity influenced by the growing fluctuation of the n=0 component.Finally,detailed convergence tests are conducted to validate the numerical simulations.
基金supported by the National Magnetic Confinement Fusion Science Program of China(Nos.2022YFE03090000 and 2022YFE03060002)National Natural Science Foundation of China(No.12375214)+3 种基金China National Nuclear Corporation Fundamental Research Program(No.CNNC-JCYJ-202236)Innovation Program of Southwestern Institute of Physics(No.202301XWCX006-04)supported by Youth Science and Technology Innovation Team of Sichuan Province(No.2022JDTD0003)US DoE Office of Science(Nos.DE-FG02-95ER54309 and DE-FC02-04ER54698)。
文摘Effects of three-dimensional(3D)magnetic field perturbations due to feedback control of an unstable n=1(n is toroidal mode number)resistive wall mode(RWM)on the energetic particle(EP)losses are systematically investigated for the HL-3 tokamak.The MARS-F(Liu et al 2000 Phys.Plasmas 73681)code,facilitated by the test particle guiding center tracing module REORBIT,is utilized for the study.The RWM is found to generally produce no EP loss for cocurrent particles in HL-3.Assuming the same perturbation level at the sensor location for the close-loop system,feedback produces nearly the same loss of counter-current EPs compared to the open-loop case.Assuming however that the sensor signal is ten times smaller in the close-loop system than the open-loop counter part(reflecting the fact that the RWM is more stable with feedback),the counter-current EP loss is found significantly reduced in the former.Most of EP losses occur only for particles launched close to the plasma edge,while particles launched further away from the plasma boundary experience much less loss.The strike points of lost EPs on the HL-3 limiting surface become more scattered for particles launched closer to the plasma boundary.Taking into account the full gyro-orbit of particles while approaching the limiting surface,REORBIT finds slightly enhanced loss fraction.
基金supported by National Natural Science Foundation of China (Nos. 12205033, 12105317, 11905022 and 11975062)Dalian Youth Science and Technology Project (No. 2022RQ039)+1 种基金the Fundamental Research Funds for the Central Universities (No. 3132023192)the Young Scientists Fund of the Natural Science Foundation of Sichuan Province (No. 2023NSFSC1291)
文摘Many magnetohydrodynamic stability analyses require generation of a set of equilibria with a fixed safety factor q-profile while varying other plasma parameters.A neural network(NN)-based approach is investigated that facilitates such a process.Both multilayer perceptron(MLP)-based NN and convolutional neural network(CNN)models are trained to map the q-profile to the plasma current density J-profile,and vice versa,while satisfying the Grad–Shafranov radial force balance constraint.When the initial target models are trained,using a database of semianalytically constructed numerical equilibria,an initial CNN with one convolutional layer is found to perform better than an initial MLP model.In particular,a trained initial CNN model can also predict the q-or J-profile for experimental tokamak equilibria.The performance of both initial target models is further improved by fine-tuning the training database,i.e.by adding realistic experimental equilibria with Gaussian noise.The fine-tuned target models,referred to as fine-tuned MLP and fine-tuned CNN,well reproduce the target q-or J-profile across multiple tokamak devices.As an important application,these NN-based equilibrium profile convertors can be utilized to provide a good initial guess for iterative equilibrium solvers,where the desired input quantity is the safety factor instead of the plasma current density.
基金supported by the Key Project of Knowledge Innovation Program of Chinese Academy of Sciences (No. KJCX3.SYW.N4)
文摘In this paper, gas control on EAST in open and closed loop is discussed and its implementation into EASTPCS （plasma control system for the experimental advanced supercon- ducting tokamak） is introduced. Using a model to describe the plasma density response to the gas puff command, a gas simulation server （simserver） using MATLAB simulink tools and real time workshop was built up. Proper operation of the gas control algorithm was verified using this simserver. The simulation results suggested that the gas control can be applied in the next EAST campaign.
基金supported by National Natural Science Foundation of China(No.10835009)the Key Project of Knowledge Innovation Program of Chinese Academy of Sciences(No.KJCX3.SYW.N4)
文摘The EAST plasma control system （PCS） is in continuous development to satisfy the EAST experimental requirements. In order to realize low latency and distortion-free signal transmission between PCS and servo systems such as the poloidal field power supply, in-vessel coil power supply and real-time scope, reflective memory boards （RFM） were applied. The new hardware layout and enhanced performance are reported. Newly implemented PCS control algorithms for gas control and real-time data display are also presented.
基金supported by National Natural Science Foundation of China (No. 10725523)
文摘Reconstruction of experimental axisymmetric equilibria is an important part of tokamak data analysis. Fourier expansion is applied to reconstruct the vessel current distribution in EFIT code. Benchmarking and testing calculations are performed to evaluate and validate this algorithm. Two cases for circular and non-circular plasma discharges are presented. Fourier expansion used to fit the eddy current is a robust method and the real time EFIT can be introduced to the plasma control system in the coming campaign.
基金supported by the National Magnetic Confinement Fusion Science Program of China(Nos.2011GB102000,2012GB103000 and 2015GB103000)
文摘A long pulse electron cyclotron resonance heating（ECRH）system has been developed to meet the requirements of steady-state operation for the EAST superconducting tokamak,and the first EC wave was successfully injected into plasma during the 2015 spring campaign.The system is mainly composed of four 140 GHz gyrotron systems,4 ITER-Like transmission lines,4 independent channel launchers and corresponding power supplies,a water cooling,control ＆inter-lock system etc.Each gyrotron is expected to deliver a maximum power of 1 MW and be operated at 100-1000 s pulse lengths.The No.1 and No.2 gyrotron systems have been installed.In the initial commissioning,a series of parameters of 1 MW 1 s,900 k W 10 s,800 k W 95 s and650 k W 753 s have been demonstrated successfully on the No.1 gyrotron system based on calorimetric dummy load measurements.Significant plasma heating and MHD instability suppression effects were observed in EAST experiments.In addition,high confinement（H-mode）discharges triggered by ECRH were obtained.
基金supported by the National Magnetic Conlinement Fusion Science Program of China(Nos.2015GB102000 and 2015GB103000)
文摘In the 2016 EAST experimental campaign,a steady-state long-pulse H-mode discharge with an ITER-like tungsten divertor lasting longer than one minute has been obtained using only RF heating and current drive,through an integrated control of the wall conditioning,plasma configuration,divertor heat flux,particle exhaust,impurity management,and effective coupling of multiple RF heating and current drive sources at high injected power.The plasma current（Ip - 0.45 MA） was fully-noninductively driven（Vloop 〈 0.0 V） by a combination of-2.5 MW LHW,-0.4 MW ECH and -0.8 MW ICRF.This result demonstrates the progress of physics and technology studies on EAST,and will benefit the physics basis for steady state operation of ITER and CFETR.
基金the National Key R&D Program of China(No.2022YFE03010003)National Natural Science Foundation of China(No.12275309).
文摘In 2021,EAST realized a steady-state long pulse with a duration over 100 s and a core electron temperature over 10 keV.This is an integrated operation that resolves several key issues,including active control of wall conditioning,long-lasting fully noninductive current and divertor heat/particle flux.The fully noninductive current is driven by pure radio frequency(RF)waves with a lower hybrid current drive power of 2.5 MW and electron cyclotron resonance heating of 1.4 MW.This is an excellent experimental platform on the timescale of hundreds of seconds for studying multiscale instabilities,electron-dominant transport and particle recycling(plasma-wall interactions)under weak collisionality.
基金supported by National Natural Science Foundation of China (No.10835009)the China 973 project (No. 2009GB103000)
文摘A new method, by using eigenmodes to reduce the fitting parameters and precalculated eddy current based on a lump parameter circuit equation, is applied to reconstruct the vacuum field for EAST plasma startup.
基金supported by National Natural Science Foundation of China (No. 10375031)
文摘Magnetohydrodynamic （MHD） n=1 kink mode with n the toroidal mode number is studied and the operational beta limit, constrained by the mode, is calculated for the equilibrium of HL-2A by using the GATO code. Approximately the same beta limit is obtained for configurations with a value of the axial safety factor q0 both larger and less than 1. Without the stabilization of the conducting wall, the beta limit is found to be 0.821% corresponding to a normalized beta value of βN^c=2.56 for a typical HL-2A discharge with a plasma current Ip=0.245 MA, and the scaling of βN^c -constant is confirmed.
基金Project supported by National Natural Science Foundation of China (Grant Nos 10725523 and 10835009)
文摘The experimental advanced superconducting tokamak （EAST） is the first full superconducting tokamak with a D-shaped cross-sectional plasma presently in operation. Its poloidal coils are relatively far from the plasma due to the necessary thermal isolation from the superconducting magnets, which leads to relatively weaker coupling between plasma and poloidal field. This may cause more difficulties in controlling the vertical instability by using the poloidal coils. The measured growth rates of vertical stability are compared with theoretical calculations, based on a rigid plasma model. Poloidal beta and internal inductance are varied to investigate their effects on the stability margin by changing the values of parameters αn and γn（Howl et al 1992 Phys. Fluids B 4 1724）, with plasma shape fixed to be a configuration with k = 1.9 and 5 = 0.5. A number of ways of studying the stability margin are investigated. Among them, changing the values of parameters κ and li is shown to be the most effective way to increase the stability margin. Finally, a guideline of stability margin Ms（ki, li, A） to a new discharge scenario showing whether plasmas can be stabilized is also presented in this paper.
基金National Natural Science Foundation of China (No.10775137)the JSPS-CAS Core-University Program in the field of Plasma and Nuclear Fusion
文摘Lower hybrid simulation code （LSC） is integrated into transport code ONETWO to model discharges of EAST with heating and current drive using both lower hybrid wave and neutral beam injection. To test the integration, the current profiles driven by both lower hybrid wave and neutral beam injection are presented for a given equilibrium and sample density and temperature profiles before carrying out a time advancing simulation. For a steady state H-mode discharge, the temperature profiles are evolved using GLF23 transport model with a boundary value specified at the normalized minor radius pN = 0.9. A case of steady state with non-inductively driven current accounting for up to 80% is presented.
文摘Blast-wave-driven hydrodynamic instabilities are studied in the presence of a background B-field through experiments and simulations in the high-energy-density(HED)physics regime.In experiments conducted at the Laboratoire pour l’utilisation des lasers intenses(LULI),a laserdriven shock-tube platform was used to generate a hydrodynamically unstable interface with a prescribed sinusoidal surface perturbation,and short-pulse x-ray radiography was used to characterize the instability growth with and without a 10-T B-field.The LULI experiments were modeled in FLASH using resistive and ideal magnetohydrodynamics(MHD),and comparing the experiments and simulations suggests that the Spitzer model implemented in FLASH is necessary and sufficient for modeling these planar systems.These results suggest insufficient amplification of the seed B-field,due to resistive diffusion,to alter the hydrodynamic behavior.Although the ideal-MHD simulations did not represent the experiments accurately,they suggest that similar HED systems with dynamic plasma-β(=2μ_(0)ρv^(2)/B^(2))values of less than∼100 can reduce the growth of blast-wave-driven Rayleigh–Taylor instabilities.These findings validate the resistive-MHD FLASH modeling that is being used to design future experiments for studying B-field effects in HED plasmas.
基金This material is based upon work supported by the Department of Energy National Nuclear Security Administration under Award Number DE-NA0001944the University of Rochester,and the New York State Energy Research and Development Authority.
文摘Targets for low-adiabat direct-drive-implosion experiments on OMEGA must meet rigorous specifications and tight tolerances on the diameter,wall thickness,wall-thickness uniformity,and presence of surface features.Of these,restrictions on the size and number of defects(bumps and depressions)on the surface are the most challenging.The properties of targets that are made using vapor-deposition and solution-based microencapsulation techniques are reviewed.Targets were characterized using confocal microscopy,bright-and dark-field microscopy,atomic force microscopy,electron microscopy,and interferometry.Each technique has merits and limitations,and a combination of these techniques is necessary to adequately characterize a target.The main limitation with the glow-discharge polymerization(GDP)method for making targets is that it produces hundreds of domes with a lateral dimension of 0.7-2 μm.Polishing these targets reduces the size of some but not all domes,but it adds scratches and grooves to the surface.Solution-made polystyrene shells lack the dome features of GDP targets but have hundreds of submicrometer-size voids throughout the wall of the target;a few of these voids can be as large as~12 μm at the surface.
基金supported by the National Magnetic Confinement Fusion Science Program of China(Nos.2012GB105000 and 2014GB103000)
文摘For the advanced tokamak,the particle deposition and thermal load on the divertor is a big challenge.By moving the strike points on divertor target plates,the position of particle deposition and thermal load can be shifted.We could adjust the Poloidal Field（PF） coil current to achieve the strike point position feedback control.Using isoflux control method,the strike point position can be controlled by controlling the X point position.On the basis of experimental data,we establish relational expressions between X point position and strike point position.Benchmark experiments are carried out to validate the correctness and robustness of the control methods.The strike point position is successfully controlled following our command in the EAST operation.
基金supported by the National Key Research and Development Program of China(Nos.2022YFE03060002,2019YFE03090100)by the Innovation Program of Southwestern Institute of Physics(No.202001XWCXRC001)partly supported by the Youth Science and Technology Innovation Team of Sichuan Province(No.2022JDTD0003)。
文摘Transport of fast ions is a crucial issue during the operation of ITER.Redistribution of neutral beam injection(NBI)fast ions by the ideal internal magnetohydrodynamic(MHD)instabilities in ITER is studied utilizing the guiding-center code ORBIT(White R B and Chance M S 1984Phys.Fluids 272455).Effects of the perturbation amplitude A of the internal kink,the perturbation frequency f of the fishbone instability,and the toroidal mode number n of the internal kink are investigated,respectively,in this work.The n=1 internal kink mode can cause NBI fast ions transporting in real space from regions of 0<s≤0.32 to 0.32<s≤0.53,where s labels the normalized plasma radial coordinate.The transport of fast ions is greater as the perturbation amplitude increases.The maximum relative change of the number of fast ions approaches 5%when the perturbation amplitude rises to 500 G.A strong transport is generated between the regions of 0<s≤0.05 and 0.05<s≤0.12 in the presence of the fishbone instability.Higher frequency results in greater transport,and the number of fast ions in 0<s≤0.05 is reduced by 30%at the fishbone frequency of 100 k Hz.Perturbations with higher n will lead to the excursion of fast ion transport regions outward along the radial direction.The loss of fast ions,however,is not affected by the internal MHD perturbation.Strong transport from 0<s≤0.05 to 0.05<s≤0.12 does not influence the plasma heating power of ITER,since the NBI fast ions are still located in the plasma core.On the other hand,the influence of fast ion transport from 0<s≤0.32 to 0.32<s≤0.53 needs further study.
