Neutral beam deposition in large,finite-beta noncircular tokamak plasmas

A parametric pencil beam model is introduced for describing the attenuation of an energetic neutral beam moving through a tokamak plasma. The nonnegligible effects of a finite beam cross-section and noncircular shifted plasma cross-sections are accounted for in a simple way by using a smoothing algorithm dependent linearly on beam radius and by including information on the plasma flux surface geometry explicitly. The model is bench-marked against more complete and more time-consuming two-dimensional Monte Carlo calculations for the case of a large D-shaped tokamak plasma with minor radiusa=120 cm and elongationb/a=1.6. Deposition profiles are compared for deuterium beam energies of 120–150 keV, central plasma densities of 8×10¹³ to 2×10¹⁴ cm^–3, and beam orientation ranging from perpendicular to tangential to the inside wall.     

Rview of pellet fueling

The injection of frozen pellets composed of the isotopes of hydrogen has become the leading candidate for refueling fusion power reactors based on the tokamak concept. This lofty position has been reached partly as a result of efforts to find an attractive solution to the perplexing problem of depositing atoms of fuel deep within the magnetically confined, hot plasma, and because of some recent experimental successes. To some extent, the relative merits of this technique will depend upon the distance that the cryogenic pellet will penetrate such a plasma, and the early exploratory research has addressed this problem on both theoretical and experimental fronts. The conclusion from the theoretical effort is that a protective blanket consisting of hydrogenic gas or cold plasma will envelope the pellet and partially shield the surface from the intense plasma heat flux. The blanket prolongs pellet lifetime, but penetration to the plasma center might require pellet injection velocities in excess of 10 km/s. The need for central penetration has not yet been established either theoretically or experimentally. The experiments performed to date have verified the existence of a shielding mechanism in general, and pellet ablation models that incorporate neutral gas shielding in particular are in adequate agreement with the experiments. Magnetic shielding effects are expected to contribute to, but not dominate, self-shielding in the higher plasma temperature regimes of the future. The tokamak plasma has demonstrated a surprising resilience even to massive density perturbations caused by the large refueling pellets used in present experiments. The characteristic discharge behavior is qualitatively not unlike that observed with gas puffing; but, for the first time, central plasma fueling has been studied, and this does not appear to be superior to refueling by partial pellet penetration. If relatively large pellets containing a significant fraction of the total plasma charge are acceptable in the present resistive plasma regimes, then it can be argued that they should have little impact on the gross stability of a hot thermonuclear tokamak plasma. Large pellets are preferable from the standpoint of attaining deep penetration, and this has important implications for the technology of pellet injection. The interesting velocity regime of 1 km/s has already been achieved with simple gun-type devices and this should be adequate for near-term tokamak experiments. Further improvements are anticipated, but the 10 km/s and above regime is uncertain; and, if current theory and experiments extrapolate to the future, such velocities might be desirable but unnecessary.     

托卡马克超声分子束注入加料的进展

超声分子束注入作为一种新的托卡马克加料方法由作者在1992年首次提出并于当年在中国环流器一号(HL-1)装置演示成功,随后相继应用于中国环流器新一号(HL-1M)和中国科学院超导托卡马克HT-7装置。超声分子束注入等离子体呈现出电子密度峰化和温度中空分布的特征;等离子体流极向旋转速度提高,边缘扰动被抑制,等离子体能量约束得到改善。加料效率较常规脉冲送气提高一倍,而滞留器壁的粒子大为减少。近期开展的高气压氢超分子束注入实验,在束流中发现团簇流,可注入等离子体中心区域。多脉冲分子束注入形成电子密度的阶跃上升,如同冰弹丸注入效果。近年来该项技术已陆续应用于国外大型托卡马克和仿星器,是核聚变装置稳态运行的一种有效的加料方法。     

超声分子束用于聚变等离子体加料

超声分子束注入在中国环流器一号和新一号装置首次采用。与常规送气相比,由于气体粒子注入深化,形成电子密度的峰化和密度极限的提高,并导致约束的改善。欧姆加热等离子体的能量约束时间的线性范围增长到ｎ＾－ｅ＝４×１０＾１９ｍ＾－３,实验结果表明,超声分子束注入是一种先进而简单的气体加料方法。     

Engineering design for the HL-2M tokamak components

The HL-2A tokamak will be modified into HL-2M. The Bt at the plasma center (major radius R = 1.78 m) is 2.2 T, the minor radius is 0.65 m. The plasma current I_P of HL-2M will reach up to 2.5 MA, the elongation and triangularity is more than 1.8 and more than 0.5, respectively. The vacuum vessel torus consists of 20 sectors with “D” shaped cross-section and double wall structure. 20 toroidal field coil bundles comprise 140 turns which are designed with demountable joints, the poloidal field coils system consists of 25 coils. The engineering design and calculation for field coil system, vacuum vessel, support structure, etc. are finished, many key issues for manufacture process have been discussed with industry and the fabrication of main components of HL-2M tokamak will be carried out in factories.     

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.     

Equilibrium evolution on the resistive time scale in a tokamak reactor

In the Engineering Test Facility (ETF), the plasma pulse duration is expected to be hundreds of seconds, which is comparable to the resistive time scale that governs the resistive diffusion of the equilibrium. The resistive evolution of the safety factorq profile may, for MHD stability reasons, limit the duration of the plasma burn in a tokamak reactor. It may be possible to control this evolution and extend the plasma burn time through proper profile tailoring. We study the evolution of theq profile on the resistive time scale numerically using a one- and-one-half-dimensional (1 1/2-D) single fluid transport code. Two high beta (T 7–16%) cases are considered: (a) a beam-driven hydrogen plasma with no nuclear alpha heating for which the beam energy is used as a device to control the temperature profile, and (b) an ignited D-T plasma in which the neutral injection has been turned off. For the beam-driven plasma, it is shown that low beam energy heating profiles lead to resistive steady states having broad temperature profiles and flatq profiles, while high beam energy heating profiles lead to resistive steady states having peaked temperature profiles and deepq profiles. The centralized nuclear heating in an ignited D-T plasma causes the evolution of theq profile for this case to behave much like that in the high energy, beam-driven case: namely, theq values near the plasma center decrease on the resistive time scale until a deep, resistive, steady-stateq profile is reached.Research sponsored by the Office of Fusion Energy, U.S. Department of Energy, under contract W-7405-eng-26 with the Union Carbide Corporation.     

Divertor Experiments with MBI and Strong Gas Puffing on HL-2A

In the HL-2A 2004 experiment campaign, pulsed molecular beam injection （MBI） and strong hydrogen gas puffing under the divertor configuration were used for gas fueling. The experimental results show that the MBI of hydrogen can reduce the heat flux to the divertor target plate. The electron temperature measured by the Langmuir probe array decreases significantly during the injection of the molecular beam whereas the electron density increases. This indicates that the plasma pressure near the target plates tends to be constant at a new equilibrium level. In the divertor plasmas with strong hydrogen gas puffing a high plasma density up to 4.4 × 10^19 m^-3 was achieved. In addition, a phenomenon similar to the partially detached divertor regime was observed, which is being studied in open divertor tokamaks such as DIII-D to reduce the peak heat flux on the target plates near the separatrix. After a strong gas puffing the electron temperature measured on the outer divertor target plate near the separatrix decreases till below 5 eV or even lower, but that of the farther outer divertor target plate does not change obviously; and the CIII and the Ha emissions at the plasma edge decrease as expected, but the Ha emission near the Xpoint increases. These results reflects some interesting characteristics, which needs to be studied by further modeling and experiments.     

Optic diagnosis of neutral beam injection on HL—1M

During the operation of a high-power neutral beam injection(NBI)system on the HL-1M tokamak,an optical diagnostic means using CCD camera was developed to characterize the NBI performance.The vacuum valve opening process and NBI period in the HL-1M experiment were displayed by a lot of photos taken with this means.Thus,the Hα emission profiles of the neutral beam(NB) and its interaction with plasma were given.Finally,the reason possible for plasma breakdown during NB1 model II discharge was investigated.this in-situ diagnosis can provide more information of the NB1.     

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.     

Experimental Observation of the Edge and Divertor Plasma in HL-2A

Edge plasma characteristics were studied by a fast-scanning 4-probe array and a Much/Reynolds stress/Langmuir 10-probe movable array in the boundary region. These probes could measure the edge plasma temperature, density, poloidal electric field, radial electric field, Reynolds stress, poloidal rotation velocities and their profiles, which could be obtained by changing the radial positions of the probe array shot by shot. The measured results were used to analyse plasma confinement, turbulent fluctuations and correlations. The fixed flush 3-probe arrays were mounted on the 4-divertor neutralization plates at the same toroidal cross-section in the divertor chamber. These probes were used to measure the profiles of the electron temperature, density and float potential in the divertor chamber. Edge plasma behaviours in both limiter configuration and divertor configuration are compared. The decay lengths of the edge temperature and density were measured and is emphasized for plasma behaviours of the supersonic molecular beam injection and lower hybrid current drive. The dependence of the radial gradient of Reynolds stress on the poloidal flow and the radial gradient of the electric field on turbulent loss are discussed.     

The Principle of Plasma Boundary Identification on HL-2A Tokamak

Plasma boundary identification is a basic task for studies on equilibrium and confinement in a divertor tokamak. With the progress on the experiments after engineering experiments, the boundary identification becomes an important issue for HL-2A. In order to satisfy the requirements of preciseness, simplified measurements and quickness, the filament current method instead of solving the equilibrium equations is used to identify plasma boundary on HL-2A. The involved principle, mathematics and the progresses, which have been made with this method, are given.     

Optical path design of phase contrast imaging on HL-2A tokamak

A phase contrast imaging(PCI) diagnostic has recently been developed on HL-2 A tokamak. It can diagnose plasma density fluctuations with maximum wave number of 15 cm~(-1) and wave number resolution of 2 cm~(-1). The time resolution reaches 2 μs. A 10.6 μm CO_2 laser is expanded to a beam with a diameter of 30 mm and injected into the plasma as an incident beam,injecting into plasma. The emerging scattered and unscattered beams are contrasted by a phase plate. The ideas of optical path design are presented in this paper, together with the parameters of the main optical components. The whole optical path of PCI is not only carefully designed, but also constructed on HL-2 A. First calibration results show the ability of this system to catch plasma turbulence in a wide frequency domain.     

Investigation on the Fatigue Characteristic in Support Structure of HL-2M Tokamak

In order to obtain enhanced plasma parameters a complete new tokamak HL-2M is now under construction in Southwestern Institute of Physics. To assure the structural safety of the device for the entire operation cycle, one of the most important issues is the lifetime-limiting effects due to the pulsed operation mode. Fatigue is one of the major failure modes to be considered in mechanical design, and pulsed operation imposes stress with significant alternating components on the support structure （SS）. Therefore, the reliability of the whole device is strongly affected by the stress and fatigue characteristic of the SS as the interface structure. This article introduces the SS design and details the fatigue life calculation methods based on the different characteristics of the sub-structures. The fatigue life in hazardous areas of the toroidal field coils anti-torque structure （TFCs-ATs） has been determined by non-linear analysis results. And with the stress- time history data of the vacuum vessel ＆ poloidal field coils support structure （VV＆PFCs SS）, the fatigue analysis of the hot spots has been completed based on rain-flow counting method and linear cumulative damage method. The calculated minimum fatigue life on TFCs-ATs and VVSzPFCs SS is 4.743E＋05 and 1.805E＋06 cycles, respectively. And the calculated fatigue life on sub-structures can meet the required life for HL-2M tokamak： 1.0E＋05 cycles.     

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.     

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.     

Development of the pellet injection system on the J-TEXT tokamak

Pellet injection is an attractive technology for core-fueling and magnetohydrodynamic study in magnetic-confinement fusion devices like tokamaks and stellarators. It can inject solid hydrogen/deuterium pellets into the plasma with deeper density deposition compared with other fueling methods, such as gas puffing. A three-barrel H₂ pellet injection system was installed on the J-TEXT tokamak and experiments were carried out. The pellets are formed in three barrels cooled by a cryocooler and compressor system at around 9 K, and are 0.8 mm/1 mm diameter and 0.8 mm length. The pellet is launched by helium propellant gas and injected from the low-field side of the plasma. The normal range of pellet speed is 210–310 m s⁻¹ for different propellant gas pressures. Due to the three-barrel structure, the number of injected pellets can be adjusted between one and three. Pellets can be launched sequentially with arbitrary time intervals, which enables flexible applications. The results of the experiments show that pellet fueling efficiency can reach 50%. The energy confinement time increased by about 7.5‒10 ms after pellet injection.     

Fuelling and Diagnostics with Pellet Injection in the HL-1M Tokamak

The pellet injection experiments for fuelling and diagnostics have been carried out on the HL-1M tokamak. The eight-pellet injector was installed on HL-1M. A reliable monitordetector and camera system was set up to take initial pellet photographs and measure the initial pellet speed and size. High fuelling efficiency of 60 % - 100 % and a density profile with a peaking factor of 1.8 - 2.0 were obtained. The maximum density close to 10^14/cm^3 in HL-1M was achieved with newly optimized combined fuelling techniques. Two typical models of pellet ablation have been utilized for simulative calculation of the ablation rates in HL-1M. In comparison with the distribution of the measured Hα emission intensity from the digital data of the CCD camera, the experimental result seems more optimistic for core fuelling than theoretical predictions by the two models. The safety factor profile q（r） has been extracted from the information provided by the CCD camera during the pellet injection. The reliability of the measured results depends mainly on the calibration of the imaging space position. Based on the calibration, the measured q-profile becomes more reasonable than those published previously for the same shot number and same photograph.     

Research on the phase adjustment method for dispersion interferometer on HL-2A tokamak

A synchronous demodulation system is proposed and deployed for CO_2 dispersion interferometer on HL-2 A,which aims at high plasma density measurements and real-time feedback control.In order to make sure that the demodulator and the interferometer signal are synchronous in phase,a phase adjustment(PA) method has been developed for the demodulation system.The method takes advantages of the field programmable gate array parallel and pipeline process capabilities to carry out high performance and low latency PA.Some experimental results presented show that the PA method is crucial to the synchronous demodulation system and reliable to follow the fast change of the electron density.The system can measure the lineintegrated density with a high precision of 2.0?×?10~(18)m~(-2).     

Review of the experiments for energetic particle physics on HL-2A

This paper reviews the energetic particle(EP) experiments during electron cyclotron resonance heating(ECRH) and neutral beam injection in the HL-2 A tokamak.A number of important results are summarized,which relate to ITER physics,including the behavior of the multi-mode instability,the nonlinear interaction between wave–wave and wave–particles,the losses of EP induced by the instabilities,the effect of the EP instabilities on the thermal plasma confinement and the control of the EP instabilities by means of ECRH.Systematic experiments indicate that when the drive is great enough,the nonlinear effects and the multi-mode coexistence may play an important role,which affect the transport both of the EPs and the background plasma confinement,and these instabilities could be controlled.Some new phenomena about the EP induced instabilities discovered recently on the device,such as high frequency reversed shear Alfvén eigenmodes,Alfvénic ion temperature gradient modes,the geodesic acoustic mode induced by energetic electrons excited by interaction between tearing mode and beta induced Alfvén eigenmode and double e-fishbone in negative magnetic shear discharges etc,have also been presented in the paper.