Plasma Confinement Properties Research Articles

Progress towards high performance plasmas in the National Spherical Torus Experiment (NSTX)

The major objective of the National Spherical Torus Experiment (NSTX) is to understand basic toroidal confinement physics at low aspect ratio and high βT in order to advance the spherical torus (ST) concept. In order to do this, NSTX utilizes up to 7.5 MW of neutral beam injection, up to 6 MW of high harmonic fast waves (HHFWs), and it operates with plasma currents up to 1.5 MA and elongations of up to 2.6 at a toroidal field up to 0.45 T. New facility, and diagnostic and modelling capabilities developed over the past two years have enabled the NSTX research team to make significant progress towards establishing this physics basis for future ST devices. Improvements in plasma control have led to more routine operation at high elongation and high βT (up to ∼40%) lasting for many energy confinement times. βT can be limited by either internal or external modes. The installation of an active error field (EF) correction coil pair has expanded the operating regime at low density and has allowed for initial resonant EF amplification experiments. The determination of the confinement and transport properties of NSTX plasmas has benefitted greatly from the implementation of higher spatial resolution kinetic diagnostics. The parametric variation of confinement is similar to that at conventional aspect ratio but with values enhanced relative to those determined from conventional aspect ratio scalings and with a BT dependence. The transport is highly dependent on details of both the flow and magnetic shear. Core turbulence was measured for the first time in an ST through correlation reflectometry. Non-inductive start-up has been explored using PF-only and transient co-axial helicity injection techniques, resulting in up to 140 kA of toroidal current generated by the latter technique. Calculated bootstrap and beam-driven currents have sustained up to 60% of the flat-top plasma current in NBI discharges. Studies of HHFW absorption have indicated parametric decay of the wave and associated edge thermal ion heating. Energetic particle modes, most notably toroidal Alfvén eigenmodes and fishbone-like modes result in fast particle losses, and these instabilities may affect fast ion confinement on devices such as ITER. Finally, a variety of techniques has been developed for fuelling and power and particle control.

Extension and characteristics of an ECRH plasma in LHD

One of the main objectives of LHD is to extend the plasma confinement database for helical systems and to demonstrate such extended plasma confinement properties to be sustained in the steady state. Among the various plasma parameter regimes, the study of confinement properties in the collisionless regime is of particular importance. Electron cyclotron resonance heating (ECRH) has been extensively used for these confinement studies of LHD plasma from the initial operation. The system optimizations including the modification of the transmission and antenna system are performed with special emphasis on the local heating properties. As a result, a central electron temperature of more than 10 keV with an electron density of 0.6 × 1019 m−3 is achieved near the magnetic axis. The electron temperature profile is characterized by a steep gradient similar to those of an internal transport barrier observed in tokamaks and stellarators. The 168 GHz ECRH system demonstrated efficient heating at densities more than 1.0 × 1020 m−3. The continuous wave ECRH system is successfully operated to sustain a 756 s discharge.

高周波による核融合プラズマ制御の進展 2.電子サイクロトロン波による局所加熱•電流駆動を利用した分布制御

Methods of local profile control of electron temperature and current in fusion-oriented devices using electron cyclotron (EC) waves are reviewed. Recent progress in the development of EC heating systems that enable such control is briefly described. The advantageous features of EC wave heating include local and high power density heating properties. Current drive and electron temperature profile control experiments using EC waves performed in order to improve and investigate plasma confinement properties are also discussed.

Stellarators

Stellarators are toroidal devices where the required rotational transform of the magnetic field lines is generated by external field coils and not via an induced net toroidal plasma current. This confinement scheme has the advantages that, in principle, steady-state plasma operation is possible and that it does not have to brace itself against disruptions of a toroidal plasma current. At the cost of having to give up toroidal symmetry the properties of the stellarator field can be tailored to suit reactor needs. Research focuses on the plasma confinement properties of different stellarator fields and investigates the problems arising when one extrapolates to reactor parameters.

Studies of the Confinement and the Toroidal Current Control in Heliotron J

The plasma confinement properties of Heliotron J plasmas and the toroidal current are investigated. A plasma energy of 2.5 kJ has been achieved by 70 GHz‐0.35 MW electron cyclotron heating (ECH). The energy confinement time is within the expected values as determined by the stellarator scaling law. In the high density region, however, better confinement plasmas are observed. The transition phenomena characterized by Hα signal drop are sometimes observed in such a region. The toroidal current generally affects plasma confinement since it generates a poloidal magnetic field. From this point of view, toroidal current control is studied in terms of field-component variation and electron cyclotron current drive. The zero current condition is found in the inner vertical field scan. Current control using the electron cyclotron wave is also demonstrated.

Predictive modelling of impurity seeded plasmas in JET

Seeding of impurities into JET resulted in a pronounced modification of plasma confinement properties. In order to analyse the underlying mechanisms a predictive transport modelling has been performed by means of two approaches. (i) The Chalmers advanced fluid model for temperature gradient driven instabilities is applied to determine all transport coefficients necessary for transport modelling. In the present code computations the radial profiles of the particle densities and temperatures are computed using experimentally measured sources. The computations give good agreement of the temperature profiles with the measured ones. (ii) The Jülich transport model, used earlier to simulate the plasma evolution in TEXTOR radiative improved mode, has been amended by a proper description of trapped electron transport, which is important at low collisionality typical for JET. This model incorporated into the transport code RITM, which self-consistently computes particle sources due to plasma recycling and impurity ionization, reproduces well the peaking in density and pressure profiles caused by impurity seeding.

Stellarators

Stellarators are toroidal devices where the required rotational transform of the magnetic field lines is generated by external field coils and not via an induced net toroidal plasma current. This confinement scheme has the advantages that, in principle, steady-state plasma operation is possible and that it does not have to brace itself against disruptions of a toroidal plasma current. At the cost of having to give up toroidal symmetry the properties of the stellarator field can be tailored to suit reactor needs. Research focuses on the plasma confinement properties of different stellarator fields and investigates the problems arising when one extrapolates to reactor parameters.

Characterization and conditioning of SSPX plasma facing surfaces

The Sustained Spheromak Physics Experiment (SSPX) will examine the confinement properties of spheromak plasmas sustained by DC helicity injection. Understanding the plasma-surface interactions is an important component of the experimental program since the spheromak plasma is in close contact with a stabilizing wall (flux conserver) and is maintained by a high current discharge in the coaxial injector region. Peak electron temperatures in the range of 400 eV are expected, so the copper plasma facing surfaces in SSPX have been coated with tungsten to minimize sputtering and plasma contamination. Here, we report on the characterization and conditioning of these surfaces used for the initial studies of spheromak formation in SSPX. The high pressure plasma-sprayed tungsten facing the SSPX plasma was characterized in situ using β-backscattering and ex situ using laboratory measurements on similarly prepared samples. Measurements showed that water can be desorbed effectively through baking while the removal rates of volatile impurity gases during glow discharge and shot conditioning indicated a large source of carbon and oxygen in the porous coating.

Stellarators (Experiment)

Stellarators are attractive toroidal confinement schemes because they are inherently steady-state, are likely to yield more benign plasmas and do not pose the danger of severe mechanical stresses associated with sudden disruptions of the toroidal plasma current. In principle, the properties of the stellarator field can be tailored to suit reactor needs, however, at the cost of having to give up toroidal symmetry. Experimental research focuses on the plasma confinement properties of different stellarator fields and investigates the problems arising when one extrapolates to reactor parameters.

Measurements of Nonlinear Energy Transfer in Turbulence in the Tokamak Fusion Test Reactor

The application of a new bispectral analysis technique to density fluctuation measurements in the core of the Tokamak Fusion Test Reactor indicates that the peak in the autopower spectrum usually lies in a region of linear stability. Large changes in the linear and nonlinear characteristics of the turbulence are observed as the plasma toroidal rotation and/or confinement properties are varied, while estimates of the turbulence-driven diffusivity varies only slightly with rotation. These observations are consistent with the operation of a global organizing property that may be related to the observation of Bohm-like scaling of ion thermal transport.

D-T operation on TFTR

The Tokamak Fusion Test Reactor (TFTR) has operated safely and routinely with tritium fuel since November 1993. Experiments conducted during this time have demonstrated improved plasma confinement properties, the first measurements of alpha heating and transport and the ICRF heating and current drive of D-T plasmas. Reactor level fusion power densities up to 2.8 MW m −3 have been attained as well as more than 10 MW of fusion power and 6.5 MJ of fusion energy per pulse. In support of these accomplishments, the TFTR equipment has met or exceeded the original design values. The toroidal magnetic field has been upgraded from 5.2 to 6.0 Tesla (on-axis), the neutral beams have injected up to 40 MW with a combination of deuterium and tritium fueling and the ICRF systems have been operated at various frequencies, from 30 to 64 MHz, to support an assortment of fast wave experiments. From November 1993 through July 1996, there have been 19 725 plasma shots. These include 841 deuterium-tritium pulses which have produced a total of 4.8×10 20 D-T neutrons and 1.4 GJ of fusion energy. More than 864 kCi (almost 90 g) of tritium have been processed through the TFTR facility while constrained to a site limit (`in-process') of 50 kCi. A low tritium inventory cryogenic distillation system has been recently installed and tested to provide on-site repurification of the plasma exhaust. Maintenance and repair activities have become routine in this facility. Hundreds of calibrations, repairs and installations requiring work on tritium contaminated and activated systems have been performed. Utilizing As Low As Reasonably Achievable (ALARA) principles, the radiation exposure to workers has been comparable to that experienced during the previous deuterium-only operating phase. The total annual site boundary dose, due to all pathways (prompt neutron radiation, activated air and controlled tritium release), has been less than 3 μSv, or 3% of the design limit of 100 μSv and 0.3% of the regulatory limit of 1 mSv.

Time evolution of transport properties in JT-60U H-mode plasmas with improved core confinement

The time evolution of confinement and transport properties of ICC H-mode plasmas in JT-60U has been studied. We found the change of confinement character on a timescale much smaller than the thermal energy confinement time, , at the L - H and H - L transitions. A slow change on a timescale of is also found during the H-mode phase.

Modifications of plasma edge electric field and confinement properties by limiter biasing on the KT-5C tokamak

Experiments using a biased multiblock limiter in the KT-5C tokamak show that positive biasing is more effective than negative biasing in modifying the edge electric field, suppressing fluctuations and improving plasma confinement. The biasing effect varies with the limiter area, the toroidal magnetic field and the biasing voltage. By positive biasing, the edge profiles of the plasma potential, the electron temperature and the density become steeper, resulting in a reduced edge particle flux, an increased global particle confinement time and lower fluctuation levels of the edge plasma.

An Advanced Method for Helical Coil Windings in Fusion Devices

In helical experiment devices, the accuracy of their nonplanar coils affects plasma confinement properties significantly; so coil winding accuracy is one of the crucial issues for their construction. An advanced method for helical coil windings in fusion devices was studied. This new method utilizes a geometrical characteristic of the trajectory of the helical coil, which enables us to improve the accuracy of coils, and is thus expected to improve the mechanical, electromagnetic and super-conducting properties of the coils. This paper presents the new concept of the winding method and practically shows the results of the R&D for the Large Helical Device (LHD). 8 refs., 7 figs.

Nonlinear electromagnetic gyrokinetic equations for rotating axisymmetric plasmas

The influence of sheared equilibrium flows on the confinement properties of tokamak plasmas is a topic of much current interest. A proper theoretical foundation for the systematic kinetic analysis of this important problem has been provided here by presenting the derivation of a set of nonlinear electromagnetic gyrokinetic equations applicable to low-frequency microinstabilities in a rotating axisymmetric plasma. The subsonic rotation velocity considered is in the direction of symmetry, with the angular rotation frequency being a function of the equilibrium magnetic flux surface. In accordance with experimental observations, the rotation profile is chosen to scale with the ion temperature. The results obtained represent the shear flow generalization of the earlier analysis by Frieman and Chen [Phys. Fluids 25, 502 (1982)], where such flows were not taken into account. In order to make it readily applicable to gyrokinetic particle simulations, this set of equations is cast in a phase-space-conserving continuity equation form.

Electrostatic modes as a diagnostic in Penning-trap experiments

A subset of the electrostatic modes of a cold cloud of electrons, a non-neutral electron plasma, trapped in a Penning trap has been observed and identified using a recent theoretical model. The detection of these modes is accomplished using electronic techniques which could apply to any ion species. The modes are observed in the low-density, low-rotation limit of the cloud where the cloud approaches a two-dimensional charged disk. We observe both axially symmetric and asymmetric drumhead modes. The shape, rotation frequency, and density of the cloud are found in a real-time nondestructive manner by measuring the frequency of these modes. In addition, it is found that radio-frequency sideband cooling compresses the cloud, increasing its density. The ability to measure and control the density of a trapped ion cloud might be useful for experiments on low-temperature ion\char21{}neutral-atom collisions, recombination rates, and studies of the confinement properties of non-neutral plasmas.

The use of neutral beam heating to produce high performance fusion plasmas, including the injection of tritium beams into the Joint European Torus (JET)*

The neutral beam injection (NBI) system of the Joint European Torus (JET) [Plasma Physics and Controlled Nuclear Fusion Research (International Atomic Energy Agency, Vienna, 1985), Vol. 1, p. 11] has proved to be an extremely effective and flexible heating method capable of producing high performance plasmas and performing a wide range of related physics experiments. High fusion performance deuterium plasmas have been obtained in the hot-ion (HI) H-mode regime, using the central particle fueling and ion heating capabilities of the NBI system in low target density plasmas, and in the pellet enhanced plasma (PEP) H-mode regime, where the good central confinement properties of pellet fueled plasmas are exploited by additional heating and fueling as well as the transition to H mode. The HI H-mode configuration was used for the First Tritium Experiment (FTE) in JET in which NBI was used to heat the plasma using 14 D0 beams and, for the first time, to inject T0 using the two remaining beams. These plasmas had a peak fusion power of 1.7 MW from deuterium–tritium (D–T) fusion reactions. The capability for injection of a variety of beam species (H0, D0, 3He0, and 4He0) has allowed the study of confinement variation with atomic mass and the simulation of α-particle transport. Additionally, the use of the NBI system has permitted an investigation of the plasma behavior near the toroidal β limit over a wide range of toroidal field strengths.

Magnetic perturbation effects on boundary plasmas during high power lower hybrid current drive in Tore Supra

Small time-independent magnetic perturbations (δ b r ), produced with the Tore Supra ergodic divertor coils, have been used to control thermal loads on plasma facing components, current density profiles, the transport of non-Maxwellian particles, and the confinement properties of thermal plasmas during high power ( P LH ≤3.3 MW) lower hybrid current drive (LHCD) discharges. MARFEs with 0.12 ≤ϱ m =π a 2 n e20 > I p −1 ≤0.22 (i.e., roughly a factor of 3 less than the smallest value of ϱ m previously reported) are obtained during the δ b r pulse when P LH >2.0 MW and the edge safety factor is slightly less than 3. These MARFEs generally appear to have the same characteristics as high ϱ m MARFEs and are positionally stable throughout the LHCD+δ b r pulse. Steady state conditions in which more than 90% of the total input power is radiated from a 0.15 m wide region near the high-field side wall were obtained.

Control of edge electric field by limiter biasing and electron beam injection in HYBTOK-II tokamak

Experimental research on the modification of the edge radial electric field by limiter biasing and electron beam injection from a hot cathode has been performed in a small research tokamak, HYBTOK-II. Control of the amplitude and/or the spatial profile of the radial electric field might enable us to improve the plasma confinement properties. In HYBTOK-II a bias voltage of negative or positive polarity with respect to the wall is applied to the poloidal circular limiters. In the positive biasing the plasma potential in the edge region rises in proportion to the applied bias voltage. Penetration of the radial electric field into the core region is about 2 cm, which corresponds roughly to the ion poloidal Larmor radius in the edge. The radial electric field of about 25 V/cm was generated at the edge region by the positive biasing of 50 V, and substantial suppression of the electrostatic fluctuations was observed in both the edge and scrape-off regions. By reducing the plasma potential and making the radial electric field weak in the case of the negative biasing, the electrostatic fluctuation increases significantly.

Physics basis for compact ignition experiments

Compact ignition experiments can be designed to attain peak densities n0 ≃ 1021 m-3 reliably and to achieve "low temperature" D-T ignition under conditions where the heating power of the α-particles does not exceed twice the ohmic heating power. An average current density near 1 kA/cm2 with a maximum plasma current Ip ≃ 12 MA are the reference design parameters of the representative Ignitor Ult machine. The peak density value is consistent with that of B/R0, as R0 ≃ 1.3m and BΥ ≃ 13 T, when compared to the Alcator, FT, and TFTR machines. The vacuum toroidal field BΥ is reinforced by the contribution of a large paramagnetic poloidal current (Iθ < 10 MA). The high values of the poloidal field Bp (≃3.9T) produce a strong rate of ohmic heating, while the correspoinding high values of Ip ensure that most of the fusion α-particles can be confined to deposit their energy in the central part of the plasma column. The confinement parameter n0 τE is expected to exceed ≃4 x 1020sec/m3 by a considerable margin on the basis of the confinement properties of plasmas with prevalent ohmic heating. The necessary degree of plasma purity (Zeff ≲ 1.6) is expected to be ensured by the high particle densities, the high magnetic fields and the relatively low values of the thermal loading on the first wall. Since ignition is achieved through transient conditions rather than, as is frequently assumed, a sequence of steady state conditions, numerical simulations have shown that the ohmic power can remain considerable up to ignition, when programming the rise of the plasma current and the particle density while gradually increasing the cross section of the plasma. The volume where the magnetic parameter q ⩽ 1 and macroscopic m0 = 1 plasma modes may be excited, has been shown to maintain quite low values until ignition is attained. The stability margin against these modes is made uniquely favorable by the low values of β-poloidal that are characteristic of Ignitor. An injected heating source of 16MW of ICRF power is included to broaden the scenarios under which ignition is achieved.