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Plasma experiments


Plasma experiments

Plasma experiments

In this program, research supervision focused on fusion relevant plasma based on experiments using a large fusion experiment device and basic plasma based on experiments using small devices is provided .

LHD:Large Helical Device

The LHD is a fusion experiment device in the National Institute for Fusion Science. A donut-shaped plasma (represented by light purple in the figure) is confined by a magnetic field produced by the helical coils (purple) and the poloidal coils (yellow), both of which are made of superconductors with a maximum field strength of about 3 tesla. The helical coils consist of two coils, while the poloidal coils consist of six coils with three each at the top and the bottom of the torus, as shown in the figure. The major and (averaged) minor radii of the device is 3.5 to 4.1 m and ~0.6 m, respectively. It is the largest device in the world as a magnetically confined device with a helical type magnetic field. The plasma is heated up to one hundred million degrees Celsius at maximum, using a neutral beam injection (NBI), electron cyclotron resonance heating (ECRH), and ion cyclotron range of frequency (ICRF) heating. Hydrogen, as a fuel gas of the plasma, is supplied by an iced pellet injection or by a gas puff system. Various diagnostic systems are installed in the device, as described below.

Plasma Physics and Control

In order to realize a fusion reactor in a magnetically confined device, it is necessary to maintain a high temperature (~ one hundred million degree Celsius)and a high density (~1020 m-3) plasma for more than one second. In this course, we study key issues required to control the fuel, and others. The research issues consist of production of a fusion relevant plasma, control of temperature, density and impurity content in the plasma, optimization of magnetic field structure, injection/pumping of plasma particles, and plasma-wall interaction, which are based upon physical understanding of plasma equilibrium/instability, transport processes, and balances of energy and particles. We study these issues aiming at developing an integrated control system for the fusion plasma.

(a)Production of high density and high temperature plasmas

Helical devices have an advantage over tokamaks in a high density operation. This is because the helical device not require a toroidal current and is thus free from a disruptive instability (disruption).  In tokamaks, on the contrary, the toroidal current is a prerequisite for creating the confinement magnetic field, and an achievable operating density is found to be related to the magnitude of the current.  A high density operation can mitigate an excessive rosion of the plasma facing components of the devices, and thus eases engineering constraints on the plasma facing material.  Also the high density operation can mask a detrimental effect of the neo-classical transport that is specific to the helical devices.  In these contexts, explorations of high performance operation with the high density toroidal current-free plasma is one of the most important issue for establishing high density operation scenarios in helical reactors.

The recent experiments in the LHD have demonstrated that a combination of efficient pumping and effective core fueling realizes an internal diffusion barrier, which results in a super dense core plasma of more than 1021 m-3.  This achievement opens a new operating scenario for a high density reactor operation. At the same time, new research topics emerge, such as a physics mechanism of the internal diffusion barrier, applicability of the barrier to the reactor condition, and an operating space (limit) of super dense plasmas.

It is also important to achieve a high ion temperature for realization of the nuclear fusion condition.  In the LHD, a central ion temparature as high as 94 million degrees Celsius has been successfully obtained.  It has been found that in such a high ion temperature operation the plasma rotation profile changes significantly; a relationship between the rotation profiles and energy transport is currently being investigated.  During the high ion temperature operation, an interesting phenomenon, the reduction of carbon impurity content in the core region, has also been observed.

(b) Research on magnetic field equilibrium and magnetohydrodynamic (MHD) instabilities

The nuclear fusion plasmas are confined by a “basket” of the magnetic field (equilibrium magnetic field). In order to obtain a high plasma confinement or a good MHD stability, the equilibrium field has to be optimized. In addition, an optimization of the equilibrium is explored also for realizing a high beta (β) value (a ratio of plasma pressure to magnetic field pressure). In the LHD, a maximum beta value has been updated year by year in a course of optimization of the equilibrium field by adjusting the magnetic field axis and the plasma radius, and recently a value of 5% has been achieved. In high temperature and high density plasmas, it is well known that a kind of MHD instability is driven, in which plasma escapes across the magnetic field. In the LHD, the MHD instability has been detected when the β value exceeds ~ 3%, while no remarkable deterioration in the plasma confinement is observed. The β value could then be increased further, and finally we could obtain a stable high beta operation for a duration more than 50 times the energy confinement time without any disruptive instabilities. In a theoretical study, detailed numerical simulations are being conducted for an optimum equilibrium field, under a comparative study with experiments.

The change of the magnetic field structures caused by an increase in plasma pressureThe change of the magnetic field structures caused by an increase in plasma pressure.

(c) Plasma facing materials, plasma-surface interactions and divertor plasma<

In nuclear fusion research the improvement of plasma confinement has been achieved by controlling peripheral plasmas, such as the improvements of the surface condition of plasma facing components and the creation of divertor magnetic field configuration. It is also said that one of the most important issues for realization of a nuclear fusion reactor is the wall design.

Magnetic field connection length structure of the peripheral plasma

Magnetic field connection length structure of the peripheral plasma.

In a fusion reactor, a large amount of energy and particle flux is expected to emerge from the confinement region and reach the plasma facing components. Therefore, the control of the energy and particle flux is a very important issue in order to avoid fatal damage to the plasma facing materials (the first wall and the divertor plates), together with achievement of a good pumping efficiency of Helium ash and a sufficient fuel injection, while all these issues must be compatible with a high performance core plasma. An understanding of the energy and particle transport in the peripheral plasma is a prerequisite for realizing mitigation of the excessive energy and particle loads on to the plasma facing components. For control of the plasma and for a sufficient lifetime of the first wall and the divertor plates, the following topics are also of vital importance: study of the erosion and damage of the plasma facing components; impurity production and its transport; the recycling process of fuel species, and others. The research topics related to the plasma wall interaction are evaluation of wall conditioning with a discharge cleaning and coating of the material surface, understanding and control of erosion and re-deposition processes of the first wall and the divertor plates, and others. These issues become critical especially in a steady state operation in a fusion reactor, and thus are common issues for all magnetic  confinement devices. The LHD contributes to the design and determination of the operation scenarios of next-step devices such as ITER through these research topics.

An image of high density gas bubbles created in a plasma facing component (stainless steel), photographed by a Transmission Electron Microscope (TEM).

(d) Research on plasma control

In order to realize a long pulse operation in a fusion reactor with an improved plasma confinement utilizing the super dense core plasma, the magnetic field equilibrium, the plasma density, the plasma heating power, and other conditions must be controlled for a long duration. In the LHD, a real time control of the magnetic field equilibrium during a discharge is being conducted aiming at obtaining a high beta plasma, while avoiding disruptive MHD instabilities. Such experiments can provide new insights into the realization of high beta plasma in future devices. Regarding density control, two technologies have been developed so far as a fuel supply. One is a gas puff, which emits hydrogen as a gas at room temperature. Another is a pellet injection, which shoots iced hydrogen pellets at a very high speed. The LHD is equipped with a gas puff system, which can provide a variation of puff rates 0.01 ~ 240 Pa m3/sec) with a pre-programmed or with a density feedback control system. It is also equipped with a high speed, repetitive pellet injection system. These systems contribute to particle transport study, and also enable us to achieve super dense plasma and high beta plasma through precise control of the fueling rate and efficiency.

The pellet injection system in the LHD.

Plasma Heating

Plasma heating physics and engineering development of plasma heating devices are studied. In plasma heating physics, plasma breakdown, high power plasma heating, current drive mechanism, and effects on the plasma confinement properties are investigated. In engineering development, high power neutral beam production, generation and transport of high power microwaves, and the optimization of the launcher are being developed. The LHD has adopted three types of plasma heating schemes. The first is neutral beam injection (NBI) heating. Hydrogen neutral atoms with high energy are injected into the plasma. The neutral beams are ionized in the plasma and the high energy hydrogen ions heat the bulk plasma. The second is electron cyclotron resonance heating (ECRH). The microwave injected into the plasma resonates the cyclotron motion of electrons in the plasmas and then the electrons are accelerated. And the third is ion cyclotron range of frequency (ICRF) heating. A high frequency wave is excited in the plasma, and then the ions are accelerated by the ion cyclotron resonance. These plasma heating schemes are used for current drive in the plasma as well as production and heating of high temperature plasma in the LHD.

The Large Helical Device (LHD) and installed plasma heating devices

The Large Helical Device (LHD) and installed plasma heating devices.

(a) Neutral Beam Injection (NBI)

NBI heating is a plasma heating method using high energy ions in the plasma, and it is a very powerful scheme for production of high performance plasma. The neutral beam is required so to produce energetic ions in the plasma, because the magnetic field for plasma confinement blocks the charged beam injection. In order to achieve effective neutralization of the ion beam, a positive ion source and a negative ion source are developed for relatively lower energy beam production (40 keV – 80 keV) and for high energy beam production up to 190 keV, respectively.
Three negative-ion-based beam injectors are operational in LHD, and two negative-ion sources are mounted in each beam injector. Two positive-ion-based beam injectors are also operational, and four positive ion sources are mounted in each beam injector. The capability of beam injection is 27 MW, and high performance plasma such as high β (β ~ 5.1%) plasma and high ion temperature plasma (Ti0 = 8.1 keV) have been produced by NBI heating in LHD.


Installation of NBI systems in LHD

Full-size negative ion source for LHD NBI

Research topics include transport study of NBI heated high temperature plasma, energetic ion confinement characteristics and the loss mechanism, MHD instability and anomalous loss of energetic ions induced by the MHD instability, development of high performance negative ion source, physics mechanism of negative ion production and their dynamics, RF negative ion source development for ITER, development of highly effective neutralization of energetic ion beam, and others.

(b) Electron Cyclotron Resonance Heating (ECRH)

ECRH is a plasma heating method using electron cyclotron resonance with a microwave. The high power microwave is generated with a gyrotron oscillator and is transferred almost 100m by a corrugated waveguide. Antenna systems are installed in the vacuum vessel of the LHD and control the microwave injection direction. The microwave launched from the antenna system has a Gaussian shape with a diameter of 5 cm at the plasma center. An electron internal transport barrier (ITB) has been formed in the core region and the high electron temperature beyond T e0=20 keV has been realized with ECRH plasmas to date. Recently, the pulse duration of gyrotron operation has become longer, and the steady state plasma discharge with the duration longer than 1 hour was also realized with ECRH.

The research topics of the ECRH group include plasma confinement properties and electron heat characteristics in the ECRH plasmas, physical mechanism of the formation of electron ITB, current driven by ECRH and control of the iota profile, electron Bernstein mode conversion and “over-dense” plasma heating, behavior of high energy electrons and their effect upon bulk plasma confinement, development of high power and long pulse gyrotron oscillators, development of microwave transport system with high transport efficiency, development of antenna systems, and others.

Electron temperature profi le of 200 million ℃ plasma

Electron temperature profi le of 200 million ℃ plasma

Electron cyclotron resonance heating device

Electron cyclotron resonance heating device

(c) Ion cyclotron range of frequency (ICRF) heating

ICRF is a heating method using radio frequency waves which have a frequency close to ion cyclotron frequency and mainly accelerates ions in the plasma. It is not suitable to heat ions with direct ion cyclotron resonance in magnetized plasmas. Therefore, the ions are heated by the minority ion heating method, higher harmonic heating method, and other methods. Ion Bernstein wave heating method through the mode conversion is also being developed, which is a very interesting and fundamental subject in plasma physics.

An improved ICRF antenna system aimed at longer- and higher-performance in steady state plasma operation has been installed in LHD. The radio frequency wave with a variety of frequencies is supplied for the study of the minority heating, the second harmonic heating of hydrogen ions, electron heating through mode conversion, the ion Bernstein wave heating, and other procedures.

The research topics of the ICRF group are confinement characteristics of ICRF heating plasma, confinement of energetic ions accelerated by ICRF, electron heating mechanism through mode conversion, long pulse plasma heating and steady state plasma operation, development of a radio frequency system with high power and steady state operation, development of high efficiency antenna system, development of feedback control of impedance matching, and others.

Ion cyclotron range of frequency (ICRF) heating device

Ion cyclotron range of frequency (ICRF) heating device


The measurement of plasma parameters is important for understanding the fusion plasma characteristics and, consequently, for realizing the fusion reactor. The LHD is equipped with a number of diagnostic tools which help us obtain various types of physical quantities of the plasma. Specifically, the essential parameters, such as temperature, density, and confinement time, are simultaneously obtained from the measurements, and a comprehensive analysis of those parameters shows us clues for the improvement of plasma performance.


Diagnostic devices installed in the LHD

Some diagnostic tools, such as the electrostatic probes, must be inserted into the plasma, for the measurement and the use of those devices is limited to the measurement of the plasma edge region where the temperature is low because those probes are damaged by the plasma and therefore perturb the plasma to be measured. Therefore, most plasma diagnostics are based on non-invasive methods which are typically categorized into two groups: passive and active methods. The observation of electromagnetic waves and particles emitted from the plasma is an example of passive methods. Active methods are represented by the measurements of plasma reactions to the laser, microwave, or particle beam injected into the plasma.

Huge focusing mirror for the Thomson scattering measurement

Huge focusing mirror for the Thomson scattering measurement

Passive diagnostics
  • Magnetic fluctuation measurement
  • Electron cyclotron emission measurement
  • Energetic particle measurement
  • Visible light spectroscopy
  • X-ray radiation measurement

The plasma diagnostics are continuously developing by novel ideas and cutting edge technologies.

Active diagnostics
  • Heavy ion beam probe measurement
  • Microwave reflectometry
  • Thomson scattering measurement
  • Collective Thomson scattering measurement
  • Far infrared laser interferometry
  • Divertor interferometry
  • Charge exchange spectroscopy
  • Diagnosis with TESPEL

(Tracer-Encapsulated Solid Pellet)

Ion temperature measurement

(a) temporal variation of the spectrum, (b) temporal variation of the central ion temperature, (c) spectrometer and grating, and (d) example of spectrum for emission lines of carbon ions

For the measurement of the plasma temperature the light radiated from the plasma can be used. Various energy (color) ranges of light are emitted from the plasma. The light is resolved into the energy spectrum by a spectrometer and a specifi c emission line which has information on the temperature is extracted. The emission line shape is broadened by the Doppler eff ect so that the plasma temperature is evaluated from the broadening width of the emission line. The ion temperature measurement is one of the most important diagnostics by which we can know the degree of our achievement toward the realization of a nuclear fusion reactor.

Imaging measurement

It is difficult to infer the plasma conditions in the deep inside area only from the information available at the outside of the device. Recent diagnostic techniques have made progress on that problem, and the measurement of the one-dimensional radial distribution of physical parameters in a plasma cross-section has been established.

The development of two-dimensional detectors such as the CCD and large-scale data processing techniques has further extended the spatial distribution measurement to the two-dimensional or imaging measurement. The result of the imaging measurement shows a still image of a plasma cross-section taken by a special camera which enables us to see various complicated facets of the fusion plasma.


VUV (vacuum ultra violet) telescope

Two-dimensional imaging of a poloidal plasma cross section by the tangential soft X-ray camera

Two-dimensional imaging of a poloidal plasma cross section by the tangential soft X-ray camera

The development of plasma diagnostics is a continuously advancing research field. It is exciting to find new phenomena which no one has ever seen with a diagnostic tool developed by oneself. Recently, applications of diagnostics developed for the fusion plasma for industrial research are active. The outcomes of the research at NIFS will also play an important role in those fields.

Basic Plasma 

The three main research activities in NIFS are: (i) Large Helical Device (LHD) experiment; (ii) Numerical simulation; and (iii) Fusion engineering. In addition, NIFS has several support devices with which various fundamental research projects considered to be useful in broad areas of plasma science are being performed as collaborative research with numerous universities in Japan.

Basic plasma experiments using a small device have the strong appeal of the plasma being visible to the naked eye. Further, researchers and graduate students may design experiments as they like, which is not generally feasible with large devices. If one wishes to pursue a professional career in fusion research, these experiences as a graduate student will be meaningful. In addition to the experience gained, it is possible to make a new discovery in basic plasma science, for which many collaboration studies are underway.

Basic plasma experiments in HYPER-I

Fig. 1 The HYPER-I device.


The HYPER-I device (Fig. 1) consists of a cylindrical vacuum chamber, which is 30 cm in diameter and 2 m in length, and a set of ten magnetic coils that produce a weakly-diverging magnetic field configuration. HYPER-I can produce high density (> 1019 cm-3) steady plasmas by electron cyclotron resonance heating with a 2.45 GHz microwave whose maximum power is 80 kW. By injecting the right-handed circularly polarized microwave from the high field side, the maximum attainable density significantly exceeds the cutoff density of ordinary waves.

Since the plasma is a collection of a large number of charged particles, various collective motions are formed by electromagnetic interactions. In fact, spontaneous formation of several large-scale flow patterns, or vortices, is frequently observed (Fig. 2).

In order to understand the formation mechanisms of those vortex structures, spatial distributions of many physical quantities (temperature, density, space potential, flow velocity, etc.) must be measured. So far, particle measurements using various probes are the main diagnostics in the HYPER-I device (Fig. 3). However, recent efforts focus on spectroscopic measurements using lasers. Laser measurement has advantages in that it gives little disturbance to the plasma and provides the absolute value of flow velocity. Moreover, information on neutral particles can also be obtained by laser measurement. Therefore, this research may change earlier views regarding vortex formation in which only the interaction between electrons and ions is taken into account. Progress in this field likely will come from the consideration of interaction between plasma and neutral flow.

Fig. 2 Various vortex structures observed in the HYPER-I device. (left) “Spiral vortex” resembling the Galaxy. (center) “Plasma hole” with a density hole resembling the eye of a typhoon. (right) “Counter E×B vortex” where the bright region rotates in counter direction to the E×B drift.

Fig. 3 Flow-velocity field of plasma hole measured with a directional Langmuir probe.