Improvement Of Plasma Confinement Research Articles

Radial electric field control by electron injection in tohoku university heliac

In this paper, we present experimental results of improvement for plasma confinement by controlling electric field with electron injection [1]. In order to generate the negative radial field, we inserted a hot cathode into the plasma and then biased negative voltage. We observed the plasma parameter changes after the hot cathode biasing: the increase of line density, the fluctuation suppression, the negative radial electric field formation, the plasma rotation, and the increase of the energy confinement time. These results suggest that the improvement of plasma confinement by the formation of negative radial electric field is achieved.

Excitation and Propagation of Modified Fluctuation in a Toroidal Plasma in KT-5C Device

Understanding the propagation of the turbulent perturbation in the tokamak edge plasma is an important issue to actively modify or control the turbulence, reduce the anomalous transport and improve plasma confinement. To realize active modification of the edge perturbation, a high dynamic output, broad-band, low-cost power amplifier is set up, and used to drive the active probes in the experiments on KT-5C Tokmak. By using small-size magnetic probes together with Langmiur probes, It is observed that the modified perturbation by the active probes with sufficiently driving power may spread with electrostatic mode, and electromagnetic mode as well.

Magnetoelectrostatic Confinement of Maxwellian and NonMaxwellian Plasmas

The unique features of the improved plasma confinement discovered earlier using automated optimization software have been explored. Reduced performance was associated with nonMaxwellian electron and ion distributions which are not electrostatically well confined. Enhanced confinement corresponds to Maxwellian electron and ion distributions.

Radio-frequency-driven radial current and plasma rotation in a tokamak

Plasma rotational shear is potentially important for controlling the formation and positioning of internal transport barriers that could stabilize tokamak microturbulence and improve plasma confinement. A new physical mechanism capable of inducing plasma rotation and rotational shear via the ion cyclotron resonance frequency (ICRF) heating of minority ion species in a tokamak has been proposed [Perkins, White, Bonoli, and Chan, Phys. Plasmas 8, 2181 (2001)]. The present work evaluates the validity of this mechanism under the realistic condition when fast ions are continuously heated and slowed down in a driven system. Ion dynamics are calculated with a Monte Carlo code in which wave-induced diffusion in velocity space is accounted for by a quasilinear operator. The code follows the drift trajectories of test particles in a tokamak geometry under the influence of given rf fields and collisions with the background plasma. When the heating geometry is such that no net toroidal wave momentum is injected, the rotational characteristics described in Perkins et al. are reproduced. A physical picture emerges which ascribes the directionality of the rotation as a consequence of finite drift orbit width. The stochastic nature of the wave-induced diffusion can result in a net toroidal torque on the bulk plasma. A number of predicted features can be experimentally tested as a validation of finite orbit drift physics and radio-frequency-induced orbit diffusion.

核融合炉の経済性と環境負荷 -その解析と今後の課題- 2.国内外における炉設計•経済性研究の現状

Fusionreactordesignandeconomicstudybasedonasteady-statetokamakhavebeenreviewed. Keyissues for designing an environmentally-attractive tokamak reactor are as follows; the improvement of plasma confinement, normalized beta value, density limit, and bootstrap current fraction from the aspect of plasma physics, and the maximum magnetic field of the toroidal field coil, neutron flux and thermal efficiency from the engineeringaspect. The results for the Cost of Electricity (COE) obtained by using various system an alysis codes have briefly reviewed. In comparison with other power plants, the feasibility of economically-competitive tokamak reactors is discussed.

Development of negative ion sources for the ITER neutral beam injector

A vacuum insulated beam source (VIBS), where all the components including the negative ion source and the acceleration grids are immersed in the vacuum, has been designed at Japan Atomic Energy Research Institute. The beam source is characterized by high gas flow conductance that reduces stripping losses of negative ions in the accelerator and improves the acceleration efficiency. The gas flow in VIBS was calculated using a 3-D Monte-Carlo code. At a typical operating pressure of 0.3 Pa in the negative ion source, the stripping loss of D − ions in VIBS is about 23%, which is about half of stripping loss in the gas insulated beam source (GIBS) that was designed in the ITER-FDR design. The acceleration efficiency is improved from 84 to 94%. To enhance the acceleration efficiency further, the possibility for reducing the operating pressure was experimentally studied. It was found the negative ion production in the KAMABOKO source adopted for ITER is dominated by surface production, therefore, the operating pressure can be reduced by improving plasma confinement. It is predicted that the operating pressure in KAMABOKO source can be reduced to 0.15 Pa, where the stripping loss and the acceleration efficiency in VIBS are improved to be 13 and 97%, respectively.

Improved plasma performance on Large Helical Device

Since the start of the Large Helical Device (LHD) experiment, various attempts have been made to achieve improved plasma performance in LHD [A. Iiyoshi et al., Nucl. Fusion 39, 1245 (1999)]. Recently, an inward-shifted configuration with a magnetic axis position Rax of 3.6 m has been found to exhibit much better plasma performance than the standard configuration with Rax of 3.75 m. A factor of 1.6 enhancement of energy confinement time was achieved over the International Stellarator Scaling 95. This configuration has been predicted to have unfavorable magnetohydrodynamic (MHD) properties, based on linear theory, even though it has significantly better particle-orbit properties, and hence lower neoclassical transport loss. However, no serious confinement degradation due to the MHD activities was observed, resolving favorably the potential conflict between stability and confinement at least up to the realized volume-averaged beta 〈β〉 of 2.4%. An improved radial profile of electron temperature was also achieved in the configuration with magnetic islands, minimized by an external perturbation coil system for the Local Island Divertor (LID). The LID has been proposed for remarkable improvement of plasma confinement like the high (H) mode in tokamaks, and the LID function was suggested in limiter experiments.

Improvement of Plasma Confinement on a Field-Reversed Configuration with Strong Mirror Field by Neutral Beam Injection

The first experiments of neutral beam (NB) injection into a field-reversed configuration (FRC) plasma have been performed at FIX (FRC Injection Experiment) device.1 The experimental results show that the configuration lifetime of a FRC have been expanded by about 200% compared with no NB injection case. These results indicate that several hundred kilo watts of NB injection saves several mega watts of global energy losses. For understanding these experimental phenomena, some numerical calculations have been performed. Numerically calculated ionization degree vs. distance along beam injection axis and beam trajectories show that the injected beam ions form dense regions around minimum │B│. These formed hot beam ions walls would play a role in improving these confinements on the edge layer. Some previous theoretical studies pointed out the existence of electrostatic effects in the FRC edge plasma confinements.6, 7 Since only several kilo watts NB injection improves FRC confinement, it seems that the injected hot beam ions encourage electrostatic effects on the edge layer.

A change in the characteristics of fluctuations of the boundary plasma parameters in the FT-2 tokamak upon transition to the improved plasma confinement regime

Processes in the boundary (near-wall) plasma of the FT-2 tokamak upon transition into an improved plasma confinement regime under the action of lower hybrid heating were studied by a refined plasma diagnostic method using multielectrode Langmuir probes. The experimental data show evidence of suppression of the correlation and coherency of fluctuations in the plasma density and the electric field strength as well as a reduction in the transverse particle transfer caused by the fluctuational drift flows.

Confinement Improvement in a Single Helical State of the Reversed Field Pinch Plasma on TPE-1RM20

In high theta discharges of the reversed field pinch (RFP) plasma on TPE-1RM20, it is observed that the soft X-ray signal ramps up when the amplitude of the magnetic fluctuations concentrates into a single helical mode (poloidal mode number, m =1, and toroidal mode number, n =7). Even in an extreme case where the fluctuation level of this single mode becomes quite large, as large as that in ordinary cases with a factor of about five, the soft X-ray signal is still sustained at a high level. This result indicates that the single helical state can play an important role in the improvement of plasma confinement in RFP. This assumption is confirmed by the results of the nonlinear magnetohydrodynamics (MHD) simulations of RFP plasma. The simulation shows that the single helical state can have a significant closed magnetic surface area. On the other hand, the large stochastic area of the magnetic field lines always appears in the case of the multihelical modes of the magnetic fluctuations. The numerical results...

INFLUENCE OF PERIPHERAL RADIAL ELECTRIC FIELD ON L-H TRANSITION IN T-10 TOKAMAK *)

In this paper the peripheral plasma behaviour during and after L-H transition on the tokamak T-10 is considered. Abrupt changes of Scrape-Off Layer (SOL) plasma parameters (ion saturation current, floating potential, electron temperature) are observed in regimes with transition to the improved plasma confinement. Such changes begin (5–150) ms prior to the L-H transition. The time delay of a transition depends on auxiliary plasma heating powerP aux and plasma currentI p (at constant toroidal magnetic field inductionB t).

Lower hybrid current drive experiments and improved performance on the HT-7 superconducting tokamak

The feedback control system to control plasma current and position on the HT-7 superconducting tokamak was greatly improved in early1998. Lower hybrid current drive (LHCD) experiments with the improved control system were performed to sustain long pulse discharges and to improve plasma confinement. Partial non-inductive current drive and full non-inductive current drive for several seconds by means of LHCD were demonstrated. It was observed that plasma confinementcould be considerably improved by LHCD. Experimental evidence suggests that this improvement during the LHCD phase could be due to the modification of the current profile in the outer region of the plasma. MHD modes (especially m = 2) seem unstable with such a current profile. The EFIT code was modified for the reconstruction of the magnetic surfaces in HT-7 and a test computation was performed.

LHD Plasma Modeling and Theoretical Analysis.

The transport/heating modeling and equilibrium/stability analysis have been carried out for LHD (Large Helical Device) plasmas. A new simulation code TOTAL (TOroidal Transport Analysis Linkage) is developed, which consists of the 3-dimensional equilibrium code VMEC including bootstrap current and 1-dimensional transport code HTRANS including helical-ripple transport determined as well as anomalous transport. This code clarified the favorable effect of bootstrap current on the neoclassical confinement in LHD. The 3-dimensional stability analysis using CAS3D code has been done and clarified the ballooning mode structure peculiar to the LHD high-beta plasmas. The 5-dimensional simulation code has been developed to analyze the NBI or ECH heating power depositions in LHD plasmas, and the particle orbit effects of high-energy particles are clarified. The plasma rotation analysis is also carried out related to the possibility of the electric-field transition and the plasma confinement improvement in LHD.

Spontaneous and driven reduced turbulence modes in the RFX reversed field pinch

The reversed field pinch (RFP) is a highly-sheared nearly force-free magnetic field configuration, which is almost entirely self-generated, via a dynamo process. A still open question is the level of turbulence entailed by the dynamo and how it could be mitigated in order to improve plasma confinement. The paper addresses the issue, first discussing the properties of the magnetic profiles of RFX in relation to other RFPs, then analysing some transient enhanced confinement modes which suggest that the improved confinement is due to a reduced dynamo action.

Spherical Tokamak for Material Research

At present spherical tokamaks are assumed to be prospective candidates for construction of thermonuclear reactors. These machines combine the advantages of spheromaks (compactness) and of tokamaks (improved plasma confinement). Such a combination allows achievement of higher plasma parameters in the presence of relative compactness and low cost of the main machine. Spherical tokamaks are also used for testing power stressed elements of the first wall and divertor under loads approaching those in experimental thermonuclear power reactors.

Low-pressure magnetron sputtering

The paper gives a short survey of the present state of art in low-pressure sputtering and self-sputtering. It shows main advantages of low pressure sputtering, particularly new physical conditions, i.e. (i) collisionless sight-of-line deposition process ; (ii) assistance of fast neutrals in the film growth ; and (iii) formation of films from ionised coating material and in the absence of an inert sputtering gas in the case of self-sputtering, under which novel materials with new physical properties can be formed. Principles of low-pressure sputtering are outlined. It is shown that there are two basic ways how to realise the low-pressure sputtering : (i) an improvement of plasma confinement in the sputtering discharge ; and (ii) an additional ionisation of the working atmosphere. Also, sputtering systems used for low-pressure sputtering, including that recently developed, are discussed in detail. A special attention is devoted to the magnetron with a grooved target and the magnetron enhanced with rf and microwave magnetoactive plasma.

Anomalous transport in tokamaks: examples of its suppression

Studies conducted towards an understanding of improved plasma confinement in small and large tokamaks are presented. In STOR-M, Ohmic H-modes can be induced by various means. The most recent finding is that injection of a high density compact torus triggers a sustained H-mode. For bias induced H-mode, direct evaluation of fluctuation induced particle flux has been made for Ohmic H-modes. The particle flux is intermittent and the average changes its sign from radially outward in L-mode to radially inward in H-mode. The direction and magnitude of the flux are predominantly determined by the phase difference between the density and potential fluctuations. Nonlinear correlation studies have also been made between two radial positions. In L-mode, the inner region behaves as a master and outer region as slave. In H-mode, the relationship is reversed.For high performance negative shear discharges in large tokamaks, the radial derivative of the Shafranov shift Δ′ is shown to have efficient stabilizing influence on the trapped electron drift mode and kinetic ballooning mode (KBM). The ion temperature gradient (ITG) mode is well stabilized at a modest α (the ballooning parameter).

Simulation of Burning Plasma Dynamics in ITER

The dynamics of burning plasma for various transient situations in International Thermonuclear Experimental Reactor (ITER) plasma have been simulated with the 1½-dimensional up-down asymmetry Tokamak Transport Simulation code. Attention is paid primarily to intrinsic plasma transport processes such as confinement improvement and changes of plasma profiles. A large excursion of fusion power is shown to take place with a small improvement of plasma confinement; e.g., an increase of the global energy confinement by a factor of 1.2 yields a fusion power excursion of ∼30% within a few seconds. Given this short timescale of the fusion power transient, any feedback control of fueling deuterium-tritium gas is difficult. The effect of plasma profile on fusion power excursion is studied by changing the particle transport denoted by the peaking parameter Cv. When the fusion power excursion is mild and slow, the feedback control is quite effective in suppressing the fusion power excursion and in shortening the duration time of the power transient. Changes of the pumping efficacy are also studied, and large excursions of fusion power are not observed because of a decrease of the fuel density itself when the pumping efficacy is increased; and helium ash accumulates in the case of a decrease of the pumping efficacy. Finally, magnetohydrodynamic sawtooth activity leads to a fusion power fluctuation of ±20%, although such activity is helpful for helium ash exhaust.

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.

Improved confinement in a high pinch parameter region of the reversed field pinch plasma

Discharges with improved plasma confinement in a high pinch parameter region ( Theta ~2) have been discovered in the TPE-1RMSO reversed field pinch experiment. These discharges become possible with careful control of the waveforms of the plasma current and the toroidal magnetic field after intensive discharge cleaning and are called the `Improved High Theta Mode' (IHTM). In these discharges the increase of electron density is distinct when compared with low Theta cases (~1.55). The electron temperature increases slightly with Theta but the ion temperature remains almost constant. Consequently, the poloidal beta increases, from 8% at Theta ~1.55 to 18% at Theta ~2.0, and the energy confinement time increases from 0.2 to 0.4 ms. The fluctuation level of the magnetic field corresponding to the dynamo activities is reduced by about a factor of 3 in IHTM, which may explain this improvement. Possible causes of IHTM are discussed. It is found that the m=1 mode rotates continuously during IHTM, while it always stops rotating just before a large soft X-ray emission crash occurs