Y. Hamada

Ballooning beta limits of dee- and bean-shaped tokamaks

Numerical studies of the beta limit in the first region of stability for n = ∞ ballooning modes in advanced-shaped tokamaks are presented. A higher beta value than corresponds to the present conventional beta scaling is expected for advanced-shaped tokamaks with sufficient triangularity/indentation. Extremely elongated configurations without appropriate triangularity do not lead to an increase in critical beta. Dee or bean shapes with sharply tipped cross-sections are more favourable for achieving high beta values than those with round-tip cross-sections having nearly the same safety factor profile. A new beta scaling for elongated ellipse, Dee- and bean-shaped tokamaks is derived.

MHD instabilities and their effects on plasma confinement in Large Helical Device plasmas

Characteristics of MHD instabilities and their impacts on plasma confinement are studied in current free plasmas of the Large Helical Device. Spontaneous L?H transition is often observed in high beta plasmas close to 2% at low toroidal fields (Bt ? 0.75?T). The stored energy starts to rise rapidly just after the transition accompanying the clear rise in the electron density but quickly saturates due to the growth of the m = 2/n = 3 mode (m and n: poloidal and toroidal mode numbers), the rational surface of which is located in the edge barrier region, and edge localized mode (ELM) like activities having fairly small amplitude but high repetition frequency. Even in low beta plasmas without L?H transitions, ELM-like activities are sometimes induced in high performance plasmas with a steep edge pressure gradient and transiently reduce the stored energy up to 10%. Energetic ion driven MHD modes such as Alfv?n eigenmodes (AEs) are studied in a very wide range of characteristic parameters (the averaged beta of energetic ions, ?b?, and the ratio of energetic ion velocity to the Alfv?n velocity, Vb?/VA), of which range includes all tokamak data. In addition to the observation of toroidicity induced AEs (TAEs), coherent magnetic fluctuations of helicity induced AEs (HAEs) have been detected for the first time in NBI heated plasmas. The transition of a core-localized TAE to a global AE (GAE) is also observed in a discharge with temporal evolution of the rotational transform profile, having a similarity to the phenomenon observed in a reversed shear tokamak. At low magnetic fields, bursting TAEs transiently induce a significant loss of energetic ions, up to 40% of injected beams, but on the other hand play an important role in triggering the formation of transport barriers in the core and edge regions.

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.

Study on in-situ carbon coating in JIPP T-IIU

The effectiveness of the in-situ carbon coating (carbonization) has been demonstrated to reduce the radiation loss by iron impurities during ICRF heating in the JIPP T-IIU tokamak. As a result of carbonization, the total radiation loss decreased down to one fifth of the RF power, which resulted in an increase in electrons and total stored energy compared with these conditions before carbonization. The thickness of the carbon layer was 300–900 A, and its toroidal uniformity was within a factor of 3, although only one anode and one gas-inlet were used. A thin carbide layer is formed between the C-film and the stainless steel substrate with carbonization at room temperature. The hydrogen concentration is 40–50 at.% in the carbon layer. Deposition of carbon was observed on window materials. The deposition rate was relatively less on electrical insulators compared to the deposition rate on metals.

Impurity pellet injection into current driven plasmas of the JIPP T-IIU tokamak

For interaction studies, impurity pellets of stainless steel and plastic carbon with a diameter of 0.5 mm and a velocity of 400 ± 100 m·s−1 have been injected into plasmas driven by fast wave current, with a sustained plasma current of 35-50 kA and an electron density of (2-5) × 1012 cm−3. The density rise is (6-8) × 1012 cm−3 for stainless steel pellets and 4 × 1012 cm−3 for plastic carbon pellets. At pellet injection, the current driven plasmas show no disruption, whereas all of the Ohmic discharges are disruptive. These phenomena are interpreted by a difference in the collision time with ablated pellets between thermal and non-thermal electrons. From measurements of the temporal evolution of the soft X-ray emission, the decay time of the injected impurity is estimated to be 25 ms. The effective charge states of the material of the injected pellets are calculated from the density rise and it is found that they are in the range of 0.8-1.5.

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.

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).