S. Kitagawa

Resonant helical divertor experiments in ohmic and auxiliary heated JIPP T-IIU plasmas

Abstract A series of initial resonant helical divertor (RHD) experiments have been carried out in ohmically and auxiliary heated JIPP T-IIU plasmas. Disruptive and MHD instabilities make the interpretation of the RHD results difficult but an apparent increase in the energy confinement time is observed when the helical magnetic perturbation is applied. This may be due to the suppression of MHD activity or to a reduction in the edge convective heat losses. Magnetic island effects have been observed on the floating potential of a Langmuir probe array and energy scrape-off layer widths have been measured with and without helical perturbations during ICRF operation. Basic pump limiter data is presented including ion temperatures and C4+ impurity profiles. Energy confinement times are reported in ohmically and NBI heated discharges.

Plasma current startup by lower hybrid waves in the JIPP T-IIU tokamak

The paper describes the characteristic behaviour of lower hybrid current startup in JIPP T-IIU. The current startup is carried out by the injection of 800 MHz lower hybrid waves into cold and low density plasmas (Te = 10 − 20 eV, e = (1−2) × 1012 cm−3 produced by electron cyclotron resonance or lower hybrid waves (LHW) only. The plasma current rises with a characteristic rise-time of τr ( 30-50 ms) and approaches a quasi-steady state value, Ipm (= 5-20 kA), whereupon 10-50 kW LHW power is injected into the torus, controlling the vertical field. The rise-time is inversely proportional to the bulk electron density, ne, and is comparable to the collision time of current carrying high energy electrons with the bulk plasmas. On the other hand, the current drive efficiency in the quasi-steady state is almost independent of e, i.e. Ipm/PLH = 0.4−0.7 AW−1 for e = (0.8−4) × 1012 cm−3. The conversion efficiency of RF energy injected into the torus is typically 5% during the current rise phase and 10% in the most efficienct case. The effects of the initial injection of ECH power and the observed parametric instabilities on the current startup are investigated from the viewpoint of seed current generation. During the rapid current rise when an appreciably negative loop voltage is observed, the bulk electrons are heated up to 150 eV. Various heating mechanisms responsible for the bulk electron heating are discussed.

Application of the intermediate frequency range fast wave to the JIPP TII-U plasma

A series of experiments has been conducted on the JIPP TII-U tokamak since 1989, using the newly constructed 130 MHz radiofrequency system. It has been predicted theoretically that the fast wave in this frequency range interacts weakly with particles. Two mechanisms of wave absorption have been identified in the experiment: electron Landau damping/transit time damping and 3rd harmonic ion cyclotron heating. The former mechanism is intimately connected with fast wave current drive and the latter can provide a new regime of plasma heating or a possible method of controlling the transport of alpha particles. It is found that the efficiency of the 3rd harmonic ion cyclotron heating is improved by using it in combination with neutral beam injection and ion cyclotron range of frequency heating. The heating efficiency obtained is as high as that of conventional heating. The experimental results are also analysed on the basis of a global wave theory which takes into account wave-particle interactions. These mechanisms of interaction are competing with each other; this will also be the case under more realistic reactor conditions.

Limiter H-mode and other improved confinement regimes with ICRF and NBI heating in JIPP T-IIU

The H-mode, an improved confinement regime, was attained recently in JIPP T-IIU high power heating experiments, in a limiter configuration without any shaping of the plasma cross-section. This H-mode is unusual because it was obtained with heating in the ion cyclotron range of frequencies (ICRF). It was also attained with combined ICRF and neutral beam injection (NBI) heating. The threshold power level obtained with ICRF alone is similar to or less than that obtained in the combined heating case. The dependence of the power threshold of the H-mode on various plasma parameters has been studied. It increases with the plasma current and is insensitive to the plasma density, and there is an optimum value of the toroidal field intensity. The power deposition profile for ICRF heating has been analysed with a ray tracing code and used to explain the observed dependence on the toroidal magnetic field. The paper also discusses a class of discharges with improved confinement observed in the same series of experiments. These discharges had a power level close to the H-mode threshold power and exhibited a marked improvement of confinement. They were, however, different from H-mode discharges in the time evolution of the profiles.

High power ICRF heating experiments on the JIPP T-IIU tokamak

In the JIPP T-IIU tokamak, a high power ICRF heating experiment has been conducted, up to an extremely high power density (~2 MW·m−3), with a total RF power of PRF = 2 MW. Great attention has initially been paid to the problem of impurities, and it has been found that (a) the adoption of low Z materials for the limiter, (b) in situ carbon coating (i.e. carbonization) and (c) adequate gas puffing synchronized to the RF pulse are very effective in suppressing radiation loss. With these methods, a remarkable reduction in metal impurities (especially in iron impurity) has been achieved; the total radiation loss has been reduced to less than 30-40% of the input power. In these reduced radiation loss plasmas, the characteristics of ICRF heated plasmas have been studied intensively. With an increase in the ICRF heating power, a deterioration of the energy confinement time has been observed, indicating quantitative agreement with the Kaye-Goldston L-mode scaling. It is shown that the so-called profile consistency, which is the leading feature in neutral beam heated plasmas, also holds in ICRF heated plasma. It has been observed that the electron temperature profile only responds to the safety factor q(a) and does not change when the deposition profile is controlled by tailoring the k1 spectrum.

An ICRF heating experiment using a phased antenna array

The paper discusses the results of a heating experiment in the ion cyclotron range of frequencies (ICRF) conducted on the JIPP T-IIU device with a five-element antenna array. This antenna system provides precisely shaped k∥, spectra which vary according to the relative phasing angle of the neighbouring antennas, Δ. In the experiment, Δ is scanned in a close pitch and it is found that the electron heating efficiency is drastically improved with Δ = π and the lowest efficiency is obtained with Δ = 0. This observation is analysed by calculating the k∥ spectra and then the power deposition profiles using a ray tracing code. It is also found that the effect of the phasing on the impurity problem is unexpectedly small. The reasons for the discrepancies between these data and theoretical expectations are also discussed.

Cryopump of the Neutral Beam Injector for JIPPT-II U

A high-powered neutral beam injector (NBI) with an electrical output of 3 MW was constructed to heat the tokamak plasma up to a few keV in the JIPPT-II U torus at the Institute of Plasma Physics, Nagoya University. A pumping speed of about 300 m3/sec was required in the cryopump to reduce reionization of the neutral beam particles and cold particle flux into the plasma. Several configurations of a Litton-array-type cryopump were studied by Monte Carlo calculation prior to building the actual cryopump. A cylindrical Litton-array-type cryopump stacked on the neutral beam line was designed to meet the space limitations of the NBI. A small Litton-array-type cryopump composed of a single cryopanel and radiation shield was first built for experimental testing which eventually led to the design of the full-scale cryopump. The cryopump in the NBI has successfully operated since May, 1986. The measured pumping speed of 300 m3/sec at 2.3×10−2 Pa and 4.2 K working temperature is close to the design value of 310 m3/sec.

Generation of suprathermal electrons during plasma current startup by lower hybrid waves in a tokamak

Suprathermal electrons which carry a seed current are generated by non-resonant parametric decay instability during the initial phase of lower-hybrid current startup in the JIPP T-IIU tokamak. From the numerical analysis it is found that parametrically excited lower-hybrid waves at the lower side-band can bridge the spectral gap between the thermal-velocity region and the low-velocity end of the pump power spectrum.

An aluminum vacuum vessel for a low radioactivity fusion device in near future D-T experiment

Abstract Use of a low radioactivity fusion device has been proposed for a near future D-T experiment in an R-tokamak design by the Institute of Plasma Physics, Nagoya University, in order to avoid the difficulties of repair and maintenance by remote handling. The radioactivity in the aluminum alloy design is smaller by a factor of 20–70 than that in the Phase 1 design (an initial design for this project) where a stainless steel vacuum vessel is employed. In this paper the mechanical, thermal and electric properties are compared to those of the stainless steel vacuum vessel. The electromagnetic performances are analyzed, and stress and buckling analysis is performed in both designs. The problems and advantages of the aluminum alloy vacuum vessel are shown in this study. In the results, the aluminum vacuum vessel is feasible as the component for the low radioactivity fusion device.

Part I. Conceptual design of the R-tokamak

Abstract For the study of the alpha particles behavior in DT plasma the ‘Reacting Plasma Project’ (R-project in short) was initiated in 1981. The design study of the ‘R-tokamak’ was performed from 1981 to 1985, and evolved through three versions. The first version is based upon the usual tokamak and is similar to TFTR. The remote maintenance and dismantling however is considered very difficult. For the reduction of the residual radiation level the design of DT tokamak with aluminum alloy was conducted. It was found that hand-on maintenance after 1000 DT shots was feasible. Another effect of the aluminum alloy is the increase of the shell effect and stabilization of the variously shaped tokamak. The 3rd version tokamak with the extremely shaped cross-section (the crescent tokamak) is proposed for tokamak improvement (higher beta, better confinement and low radioactivity).