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We briefly present the experimental results obtained using the various heating and current drive systems, such as electron cyclotron heating, fast wave, lower hybrid wave, ion cyclotron range of fr...

The effects of divertor biasing and edge ergodization on the divertor plasma have been investigated in the JFT-2M tokamak. Experimental results show; (1) The differential divertor biasing can change the in/out asymmetry of the divertor plasma. It especially changes the density on the ion side divertor plasma. The in/out electron pressure difference has a good correlation with the biasing current. (2) The unipolar divertor biasing can change the density profile of divertor plasma. The radial electric field and shear flow are the cause for this change. (3) The electron temperature of the divertor plasma in the H-mode with frequent ELMs induced by edge ergodization is lower than that of usual H-mode. That is due to the enhancement of the radial particle flux by frequent ELMs.

The characteristics of the edge localized model (ELM), also referred to as edge relaxation phenomenon, were investigated in H-mode plasmas of JFT-2M. It was found that the ELM is mainly a density fluctuation phenomenon at the edge and that the electron temperature at the edge, except near the separatrix, is not very strongly perturbed. Some experimental conditions controlling the ELM intensity are plasma density, plasma ion species, heating power and plasma current ramping. ELMs found in low density deuterium discharges are suppressed by raising the density. ELMs are more strongly pronounced in hydrogen plasmas than in deuterium plasmas. ELMs seen in hydrogen plasmas or in near marginal H-mode conditions are suppressed by increasing the heating power. ELMs are found to be suppressed by plasma current ramp-down whereas they are enhanced by current ramp-up. The MHD aspects of the ELM were investigated. Although no consistent MHD features of the ELMs were found, they sometimes trigger bursts of the m = 2 mode. The enhanced broadband magnetic fluctuation suggests a temporary return to the L-mode during the ELM event.

The effects of plasma edge modification on confinement have been investigated on the JFT-2M tokamak since 1985. In the course of investigations, a new enhanced confinement regime, limiter H-mode, was discovered in 1986. This paper is a review of investigations into edge recycling control methods which extend the H-mode operational regime. In addition to Ti-gettering, the following methods have been shown to reduce edge neutral buildup, to extend the operational regime and to improve n τ of H-mode discharges: (1) optimization of the gas-puff programming (decreased or terminated at the start of auxiliary heating), (2) a toroidally distributed gas-puff system, (3) edge heating by ECH, and (4) particle removal with a pump limiter.

In JFT-2M, impurity behaviors during the H-mode were studied by the spectroscopic method. During the H-mode without Edge-Localized Modes (ELMs), emissions of highly ionized ion lines increased with time, although emissions of less ionized ion lines decreased just after the H-mode transition and kept constant with time. It was found from time behavior of emissions of impurity ion lines and the iron-ion density profiles that not only metal impurity but also light impurity accumulated near the plasma center during the ELM-free H-mode. During the H-mode with ELMs, the impurity accumulation was relaxed by the onset of this instability compared with the ELM-free H-mode.

The coupling characteristics of the fast wave excited by a phased four-loop antenna array are described. Fast waves in the lower hybrid frequency range have the potential of generating plasma current in hot and dense plasmas. Fast waves are excited at the plasma edge by an oscillating magnetic field parallel to the toroidal direction. The parallel wavenumber of the excited fast waves is determined by the relative phase of the RF current on each antenna. The loading resistance of the antenna increases with density, but at densities above 2 × 1019 m−3 it starts to decrease with density, because of the steepening of the density gradient. The loading resistance is strongly dependent on the antenna phasing. The maximum loading due to surface wave excitation is obtained at Δ = 0 and the minimum at Δ = π. Substantial absorption of the excited fast waves is observed at the plasma centre when the antenna phasing is Δ = π. The absorption efficiency rises with decreasing phase velocity of the excited waves. The experimental results obtained in the coupling experiment are consistent with theoretical predictions.

The roles of the edge plasma are investigated with respect to impurity release, antenna loading, and H-mode transition. Metal impurity release is considered to be caused by an RF electric field near the antennas, and can be reduced by the out-of-phase operation. Antenna loading with H-mode plasmas is dependent on the antenna-separatrix distance, and can be increased for sufficiently small separation. Electron temperature and electron density in the edge are the key parameters for H-mode transition, and they are the same, independent of the additional heating methods, and of the impurity level.

Significant progress in the extension of pulse durations of powerful negative ion beams has been made to realize the neutral beam injectors for JT-60SA and ITER. In order to overcome common issues of the long-pulse production/acceleration of negative ion beams in JT-60SA and ITER, new technologies have been developed in the JT-60SA ion source and the MeV accelerator in Japan Atomic Energy Agency.As for the long-pulse production of high-current negative ions for the JT-60SA ion source, the pulse durations have been successfully increased from 30 s at 13 A on JT-60U to 100 s at 15 A by modifying the JT-60SA ion source, which satisfies the required pulse duration of 100 s and 70% of the rated beam current for JT-60SA. This progress was based on the R&D efforts for the temperature control of the plasma grid and uniform negative ion productions with the modified tent-shaped filter field configuration. Moreover, each parameter of the required beam energy, current and pulse has been achieved individually by these R&D efforts. The developed techniques are useful to design the ITER ion source because the sustainment of the caesium coverage in the large extraction area is one of the common issues between JT-60SA and ITER.As for the long-pulse acceleration of high power density beams in the MeV accelerator for ITER, the pulse duration of MeV-class negative ion beams has been extended by more than 2 orders of magnitude by modifying the extraction grid with a high cooling capability and a high transmission of negative ions. A long-pulse acceleration of 60 s has been achieved at 70 MW m−2 (683 keV, 100 A m−2) which has reached the power density of JT-60SA level of 65 MW m−2. No degradations of the voltage holding capability of the acceleration voltage and the beam optics due to the distortion of the acceleration grids have been observed in this power density level.These results are the longest pulse durations of high-current and high-power-density negative ion beams in the world.

We are constructing a diagnostic system to measure the electric field (≳100 kV/cm) of a free‐electron laser (FEL) beam when injected into the plasma of the Microwave Tokamak Experiment (MTX). The apparatus allows a crossed‐beam measurement, with 2 cm spatial resolution in the plasma, involving the FEL beam (with 140 GHz, ≊1 GW ECH pulses), a neutral helium beam, and a dye laser beam. After the laser beam pumps metastable helium atoms to higher excited states, their decay light is detected by an efficient optical system. Because of the Stark effect arising from the FEL electric field (E), a forbidden transition can be strongly induced. The intensity of emitted light resulting from the forbidden transition is proportional to E2. Because photon counting rates are estimated to be low, extra effort is made to minimize background and noise levels. It is possible that the lower E of an MTX gyrotron‐produced ECH beam with its longer duration pulses can also be measured using this method. Other applications of the...

A pump limiter is applied to the JFT-2M tokamak in order to reduce edge gas buildup. About 12% of the effective loss flux is removed by the pump limiter and the fueling efficiency is improved by 30%. In the high density regime, the gas buildup is successfully reduced and edge cooling by cold gas-puffing is prevented by the pump limiter. The pump limiter also improves the density limit and the energy confinement time. Improvement in the nτ value is about 34%. This pump limiter is shown to be compatible with the limiter H-mode.

The metal impurity origin is studied in DIVA which is a tokamak with an axisymmetric divertor. Ion sputtering is demonstrated to be the dominant process of the release of gold from the shell coated with gold. A cone-like formation due to the sputtering is found on the gold protection plate exposed to about 10000 discharge by using the scanning electron micrograph technique. The SEM pictures shows that the spatial distribution of the cone-like formations on the shell surface due to the ion sputtering is well correlated with the spatial distribution of the intensity of the Au–I line radiation. It is also demonstrated that a honeycomb structure can decrease the release of the metal impurity.

ICRF heating of impurity-free tokamak plasmas is studied in DIVA. The frequency is fixed at 25 MHz, which is a second-harmonic cyclotron frequency of deuteron with a toroidal magnetic field of 16.4 kG. The toroidal field BT and the ratio of proton-to-deuteron density p are varied over ranges of 12 ? BT ? 19 kG and 2 p 40%, respectively. The most favourable ion heating is obtained at BT = 18 kG and p = 5 ~ 10%. Ion temperature increases from 200 to 300 eV were measured by charge-exchange neutral analysis and also by spectroscopy, where 100 kW RF net power were applied. The results are consistent with mode-conversion theory.

In the last year, ohmic heating experiments were performed in JT-60 for about three months; May–June 1985 and March 1986. The major experimental effort has been concentrated on the divertor configuration in anticipation of H-mode operation of JT-60 in high power supplementary heating experiments starting in August 1986. Stable discharges with a divertor have been obtained at Ip = 1.85 MA and ne = 5.7 × 1019m−3. Radiated power from the main plasma in divertor discharges is reduced to one third of that in material limiter discharges. The electron density has a broad radial profile and the effective safety reaches qeff = 2.5. The neutral gas pressure in the divertor chamber becomes significantly high and the control of particle recycling has been attempted.

Thermal desorption spectroscopy (TDS) has been applied during discharge cleaning in the JFT-2M tokamak to measure hydrogen fluxes. The TDS carbon sample, thickness 0.13 mm, was heated to 1000°C by direct current and the temperature distribution of the sample surface measured by infrared thermography. The probe was exposed to three types of plasma: Taylor-type discharge cleaning (TDC), ECR discharge cleaning (ECR-DC), and glow discharge cleaning (GDC). The TDS spectra show peak desorption at around 800°C. The hydrogen flux, obtained by integration of the TDS spectrum, decreases exponentially in the radial direction with decay length 7.4 cm and 5.8 cm in TDC and ECR-DC, respectively. The relation between hydrogen fluxes and water vapour production was investigated. In TDC, the amount of water vapour depends more strongly on the electron temperature of the plasma than on the hydrogen flux. In ECR-DC, the production of water vapour increases approximately linearly with the hydrogen-flux. In GDC, hydrogen fluxes were measured by TDS but no water vapour could be detected in the residual gases during the discharge.

Neutral-beam power of up to 1.2 MW injected into the plasma has produced a volume-averaged β of up to 2.6% and a central beta β0 of up to 7%, due to the thermal components in the JFT-2 tokamak. In these beam-dominated discharges, the magnetohydrodynamic behaviour was studied. Four types of internal oscillations were observed: i) enhanced sawtooth oscillations with long repetition time and large sawtooth amplitude; ii) round sawtooth oscillations and/or reduced sawtooth oscillations with short repetition time and small sawtooth amplitude; iii) high-frequency oscillations without sawtooth oscillations, and iv) high-frequency oscillations with sawtooth oscillations. The measured beta values are compared with the critical ones as found from high-n ballooning-mode analysis, and the relationship between MHD behaviour and beta values is also investigated.

Diagnostic instruments for measuring the JT-60 plasma are fully developed and ready to operate. Each instrument and its fundamental characteristics are described as well as some aspects of the plasma parameters. The new features in JT-60 diagnostics are that the diagnostics system which has its own data processing system separate from the machine control computer, and the utilities for each diagnostics instrument (vacuum, electricity, water, air, liquid nitrogen) constitute one system in the total diagnostics system.

Electron cyclotron heating (ECH) and pre-ionization using a 28 GHz 200 kW gyrotron were investigated in the JFT-2 tokamak. Three modes of the wave were launched from different types of launch system, i.e. the extraordinary mode from the high field side of the torus, the ordinary mode from the low field side, and TE 02 mode from the top to examine their heating characteristics. The oblique launch of the extraordinary mode gave the largest central heating in consistent with the theoretical prediction. It was shown that the polarization of the launched wave as well as the effect of the mode conversion at the upper hybrid resonance layer contribute to the heating mechanism.

Transitions of L-mode plasmas to the H-mode have been induced by an electron cyclotron heating (ECH) pulse. The transitions occur when ECH is applied to plasmas preheated either by a neutral beam or by waves in the ion cyclotron range of frequency with power levels well below their own threshold power for the H-mode transition. The position of the electron cyclotron resonance layer has been scanned and it has been shown that edge heating rather than central heating is effective in inducing the transition to the H-mode.

The H-mode transition is accompanied by the development of an electron temperature pedestal of about 0.3 keV and a dramatic steepening of the density gradient inside the separatrix. A change of density profile from peaked to flat is obtained for both H- mode and ohmically heated plasmas, after an internal disruption. Hollow ne profiels are found in low Bt discharges (low q). These phenomena indicate a reduction of the electron energy transport at the edge of the plateau region.

The sputtering characteristics of various types of carbon limiter surfaces, e.g. pyrolitic graphite (PG), pulverized carbon film and carbon film produced by methane discharges, are tested. Arcing is only observed on the pulverized carbon surface in a stable discharge or the other surfaces in an unstable discharge. Ion sputtering is shown to be the dominant process of carbon efflux from the normal surfaces. Employing PG limiters and a carbon wall coated by rf-sputtering method, very low q plasmas with a good confinement characteristic are obtained. The coated carbon surface is pure and still not contaminated after 600 or more discharges.

A mode conversion theory of the fast magnetosonic wave is compared with the experimental results in a deuterium plasma with minority hydrogen. The direct measurements of toroidal wave length, toroidal damping length, and the increase of ion temperature show that parameter dependences of ion heating are well explained by the mode conversion theory including tunneling, reflection, and absorption of the fast wave at a two-ion hybrid resonance layer.

High-efficiency ICRF heating in DIVA is studied. At least 80% of the RF power coupled to the plasma is deposited with the bulk ions in the narrow central region, and about 70% of the deposited power is absorbed by the majority deuterium. The ion temperature increases up to three times the base temperature and thus exceeds the electron temperature.

Initial ohmic heating experiments in JT-60 were performed for a three month period of April-June 1985. A maximum plasma current of 1.6 MA was obtained for both divertor and limiter discharges. Low-q discharges of qeff=2.5 and high density discharges of 4.8*1019 m-3 in line-averaged density were obtained in the divertor configuration. In divertor discharges radiated loss from the main plasma can be kept at 20-30% of the ohmic input.

Numerical studies on the stability of kink and resistive tearing modes in a current-carrying linear stellarator are presented for various current profiles and helical fields. In the case of an l = 2 helical field, the magnetic shear vanishes and the stability diagram is given by the straight lines with ισ + ιδ = const., where ισ is the rotational transform due to the plasma current and ιδ is that due to the helical field. For the l = 2 stellarator with , the magnetohydrodynamic stability against kink and tearing modes is improved compared to that in tokamaks. When an l = 3 helical component exists, magnetic shear plays an important role in the stability properties. The stability diagrams become fairly complex; they are, however, understandable from properties of the Euler equation. It should be noted that the internal kink modes are made more unstable than in a tokamak, because of the l = 3 helical field.

The equilibrium of the plasma limited by the material limiter or the separatrix in the linear stellarator system was solved numerically by the finite difference method. The expansion and the deformation of the plasma column due to the longitudinal plasma current and the plasma pressure were obtained and were in agreement with the predicted values in the simple case where the analytical solutions were already found.

An arrangement of l = m baseball coils to form a toroidally linked Min-B configuration with large rotational transform is studied analytically and numerically. By optimizing an l = 3 system, closed magnetic isobars are obtained inside the last closed magnetic surface. This field configuration has high magnetic shear and occupies a significant portion of the volume of the torus.

Essential divertor functions — density and energy control — have been investigated in the JT-60 divertor with metal walls. Neutral pressure in the divertor chamber increases in proportion to n2e and high recycling state for particle exhaust is realized. At ne ≤ 6 × 1019m−3, wall pumping by the divertor plates is dominant in particle exhaust. However, at ne ≥ 6 × 1019m−3, particle exhaust by active divertor pumping systems becomes essential. Because of the effectiveness of the divertor, radiation loss in the main plasma is reduced to 5–10% of the absorbed power and impurity concentrations are significantly suppressed at very low level (Zeff = 1.5-2.0). In the experiments with graphite walls, short periods of H-mode phases were found in the outside X-point divertor discharges although improvement in energy confinement is limited to within 10%. In limiter discharges, combination of high vessel temperature (210–300 °C) and low current limiter discharges (1.5 MA) without gas puffing is effective as a wall conditioning in obtaining reproducible high density discharges. The maximum parameters, Ip = 3.2 MA, ne = 1.2 × 1020m−3, qeff = 2.2 at Bt = 4.8T, a Murakami parameter of 7.5, and a stored energy of 3.1 MJ have been attained. The maximum density is limited by the occurence of the MARFE. During high power neutral beam heating, enhanced carbon influx was observed possibly due to chemical sputtering. The wall pumping by the inner wall is still effective in decreasing the electron density from the high density region of ne ~ 1 × 1020m−3. An empirical energy confinement scaling of JT-60 has been drawn in terms of an offset linear function as τE = 0.19 I1.9PPabs + 0.062 a1.8p

Deuterium and hydrogen energy spectra during ICRF heating as well as nH/nD were measured with the E||B type mass-separated neutral particle energy analyser. High-energy tails of deuterium and hydrogen were observed in the dilute-minority case. From Fokker-Planck calculations of the charge-exchange spectrum and power-balance analysis it is shown that second-harmonic cyclotron damping by the majority deuterium is the dominant mechanism of ion heating in the dilute-minority case.

ICRF heating experiments were carried out in the JFT-2 tokamak with pure high-field-side excitation. Almost pure electron and ion heating regimes were made possible in the experiment so that the metal impurities were reduced. Power balance of the plasma heated by ICRF is revealed by a transport analysis of the experimental results. RF power deposition profiles of each heating regime are obtained. The results of this analysis may serve as an experimental basis for refining the present theoretical treatment of ICRF heating.

From June to October 1987, JT-60 achieved fusion product (ne(0). tau E*.Ti(0)) of 6*1019 m-3.keV.s with hydrogen plasma at plasma current of 2.8 to 3.1 MA with neutral beam power of approximately 20 MW. The central electron density of 1.3*1020 m-3 was obtained at plasma current of 3 MA with 13 approximately 20 MW neutral beam power and the confinement time reached 0.14-0.18 s. An offset linear scaling law like the Shimomura-Odajima scaling on confinement time will be able to reproduce experimental data better than that of the Goldston type scaling. With low beam energy injection approximately 40 keV, confinement degradation was found. Many short periods (0.05 approximately 0.1 s) of H-mode phase were found in outside X-point divertor discharges with NB or NB+RF(LH or IC) heating power above 16 MW. However, improvement in energy confinement time was limited to 10 %. The ballooning/interchange stability analyses were also made for the outside X-point divertor equilibrium in connection with H0-phase capability. Heating powers of 9.5 MW and 1.9 MW were obtained by LHRF, ICRF injection, respectively. In combined LHRF and NB heating, the incremental energy confinement time of 0.064 s was obtained, which is the same level of that of NB heating only. In combined NB and on-axis ICRF heating of low ne discharge, an incremental energy confinement time of 0.21 s was obtained, which is three times as long as those of NB or ICRF heating only. High energy beam ions were accelerated by ICRF in the central region of the plasma.

Divertor actions for controlling the impurity are investigated in DIVA, The heat exhaust function reduces the electron temperature of the plasma which contacting to the first wall, thereby reduces the ion-sputtering yield. Cooling of the scrape-off plasma with gas feeding further advances this situation and suppresses the contamination by the high- Z impurity. For a carbon and aluminum injection, the accumulation of impurities in the main plasma is reduced by the factor of 2–4 by the shielding action. The shielded impurities are rapidly guided into the burial chamber with a flow velocity roughly equal to that of the bulk scrape-off plasma. The back-flow of impurity ions from the burial chamber to the main plasma is verified to be small.

A pulse of methane gas and a short burst of aluminium have been injected into DIVA plasma at the point farthest from the divertor. For the same amount of injected impurity, the accumulation of injected impurity in the main plasma is reduced by a factor of 2 to 4 by the action of the divertor. The impurity shielded by the scrape-off plasma is rapidly guided into the burial chamber. The time needed for the poloidal excursion of the injected impurity ions in the scrape-off layer is roughly equal to that of the scrape-off plasma. Sweeping-out of the impurity ions diffusing from the main plasma into the burial chamber is also observed.

This paper describes the basic architecture of the optical (CAMAC) serial highway system first, and then the architecture of the newly developed byte serial U-port adapter for optical transmission (BUAO), which is a key component of the optical highway system, as well as the results of experiments conducted on the system. Finally, this paper describes the measurement system in Japan Atomic Energy Research Institute as an application of the system.

It is observed that fast waves launched into an electron cyclotron heated plasma from a phased four-strip-line antenna array couple effectively with core electrons. This coupling produces enhanced electron cyclotron emission and a drop in the loop voltage. These effects depend on the phase difference between adjacent strip lines and on the electron density. An absorption efficiency of 20% is deduced under the conditions {ital v}{sub {ital ph}}/v{sub e0}{similar to}3 and {ital f}{sub 0}/f{sub LH}(0){similar to}0.38. The experimental results agree well with predictions based on electron Landau damping of fast waves.

Summary Form only given, as follows. The MTX program has focused on ohmic operation of the tokamak (Alcator-C), construction of the free-electron laser (FEL) microwave source and quasi-optical transport system, and preparation for the first measurements of absorption. The first microwave pulses from the FEL to the tokamak were planned for the spring of 1989. Measurements of the absorption of single, high-peak-power pulses (140 GHz, 2 GW, 20 ns) will follow. Thereafter the FEL will be upgraded to generate pulses at 5 kHz for 10 ms, providing an average power 0.5-1.0 MW, and then further upgraded to 250 GHz and 2 MW average power. Several diagnostics have been prepared for the initial microwave absorption experiments. These include microwave measurements, fast X-ray and cyclotron emission from the heated electrons, Thomson scattering from the tail, and fast magnetic probe measurements. Modeling of absorption and wave refraction have been done, and a computer database has been established for rapid analysis of the experimental results and comparison with theoretical predictions. Following the upgrade to multiple-pulse operation and subsequent absorption measurements, experiments on heating and transport will be carried out. Attention will be directed to heating in peaked density profiles generated by pellet injection and to the effects of varying the power deposition profile. >

The neutral beam injector for JAERI Fusion Torus-2(JFT-2) was developed and is now in successful operation. This is one of the largest injection systems in Japan and was designed, fabricated and tested in about one year.

The role of runaway electrons in the formation of arcing on the electrically isolated target in the scrape-off layer of tokamaks is discussed. Runaway electrons offer the sufficient sheath potential to produce an arc spot and supply the circulating current to sustain an arc in the current rising phase and in the disruptive phase of tokamak discharge.

An experimental study of the metal impurity origin in DIVA is described. Three processes for the release of the metal impurities, that is, ion sputtering, evaporation and arcing have been identified. Among these processes, ion sputtering is the dominant process in the quiet phase of the discharge, that is the phase characterized by no spikes in the loop voltage and no heat flux to a specific part of the first wall. Moreover it is demonstrated that a honeycomb structure can decrease the release of the metal impurity. This paper is being published separate [1].