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The effects of toroidal rotation and toroidal field (TF) ripple on the edge pedestal structure and edge-localized-modes (ELMs) were examined in JT-60U. The amplitude of the TF ripple was changed by the installation of ferritic steel tiles (FSTs). The profile of toroidal rotation VT became less counter with FSTs, particularly in the case where co-NBI was used in a large volume configuration. In this case, the plasma pressure was raised across the whole profile of the plasma. At the plasma edge, higher pedestal temperature was obtained with the growth of pedestal width. However, the effect of FSTs became less significant in a small volume configuration. As the VT became less counter at the pedestal, ELM frequency fELM was reduced and ELM energy loss ΔWELM was increased at fixed power crossing the separatrix Psep. The observed larger ELM energy drop at VT in a relatively co-direction involves the ELM perturbations of the electron temperature profile across an ELM that extends radially more inward. In addition, the inter-ELM transport loss is reduced and the pedestal pressure pped is weakly raised. The effect of FSTs appeared clearly in the large volume configuration where pped is raised even at a given VT at the pedestal while pped is not changed by FSTs at the small volume configuration. This suggests that the reduction of the TF ripple, the magnitude of which can affect the VT profile, plays a role in the increasing of pped. For increased pped due to the use of co-NBI and the reduction of TF ripple, the spatial width of the pedestal ion temperature became greater.

The properties of a spherical conductive dust particle on a plasma-facing wall and sheath formation with the dust particle are considered using the self-consistent two-dimensional Particle-In-Cell (2D PIC) simulations. We obtained the self-consistent simulated values of the dust charge on the wall position for different radii of the dust particle and values of the wall potential and made the comparison with the 1D model. These results allow us to evaluate the one-dimensional model and clarify the effect of dust shielding by plasma.

Transient transport experiments are performed in plasmas with and without internal transport barriers (ITB) on LHD and JT-60U. The dependence of χe on the electron temperature, Te, and on the electron temperature gradient, ∇Te, is analysed with an empirical non-linear heat transport model. In plasmas without an ITB, two different types of non-linearity of the electron heat transport are observed from cold/heat pulse propagation: the χe depends on Te and ∇Te in JT-60U, while the ∇Te dependence is weak in LHD. Inside the ITB region, there is none or weak ∇Te dependence both in LHD and JT-60U. Growth of the cold pulse driven by the negative Te dependence of χe is observed inside the ITB region (LHD) and near the boundary of the ITB region (JT-60U).

Compatibility of advanced tokamak plasmas with high density and high radiation loss has been investigated in both reversed shear (RS) plasmas and high βp H-mode plasmas with a weak positive shear on JT-60U. In the RS plasmas, the operating regime is extended to high density above the Greenwald density (nGW) with high confinement (HHy2 > 1) and high radiation loss fraction (frad > 0.9) by tailoring the internal transport barriers (ITBs). With a small plasma-wall gap, the radiation loss in the main plasma (inside the magnetic separatrix) reaches 80% of the heating power due to metal impurity accumulation. However, high confinement of HHy2 = 1.2 is sustained even with such a large radiation loss in the main plasma. By neon seeding, the divertor radiation loss is enhanced from 20% to 40% of the total radiation loss. In the high βp H-mode plasmas, high confinement (HHy2 = 0.96) is maintained at high density ( ) with high radiation loss fraction (frad ~ 1) by utilizing high-field-side pellets and argon (Ar) injection. The high is attributed to the formation of strong density ITB. Strong core-edge parameter linkage for confinement improvement is observed, where the pedestal pressure and the core plasma confinement increase together. The measured radiation profile including contributions from all impurities in the main plasma is peaked, and the central radiation is ascribed to the contribution from Ar accumulated inside the ITB. Impurity transport analyses indicate that the Ar density profile, twice as peaked as the electron density profile, which is the same level as that observed in the high βp H-mode plasma, can yield an acceptable radiation profile even with a peaked density profile in a fusion reactor.

The dependence of the thermal diffusivity on the heat flux has been investigated in JT-60U plasmas by varying the neutral beam power. In positive magnetic shear (PS) plasmas, the ion thermal diffusivity (χi) in the core region generally increases with the heating power, as with the L mode at low heating powers. However, as a result of the intensive central heating, a weak internal transport barrier (ITB) is formed, and the χi value in the core region starts to decrease. Corresponding to a further increase in the heating power, a strong ITB is formed and the χi value is reduced substantially. In the case of reversed magnetic shear (RS) plasmas, on the other hand, no power degradation of the χi value is observed in any of the heating regimes. The electron thermal diffusivity (χe) is correlated with the χi value in PS and RS plasmas. Furthermore, the dependence of the ion thermal diffusivity on the radial electric field (Er) shear has also been investigated. By considering the non-locality in the relation between the Er shear and the χi value, a threshold in the effective Er shear to change the state from a weak to a strong ITB is identified.

The relationship between particle and heat transport in an internal transport barrier (ITB) has been systematically investigated in reversed shear (RS) and high βp mode plasmas of JT-60U. The electron effective diffusivity is well correlated with the ion thermal diffusivity in the ITB region. The ratio of particle flux to electron heat flux, calculated on the basis of the linear stability analysis, shows a similar tendency to an experiment in the RS plasma with a strong ITB. However, the calculated ratio of ion anomalous heat flux to electron heat flux is smaller than the experiment in the ITB region. Helium and carbon are not accumulated inside the ITB even with ion heat transport close to a neoclassical level, but argon is accumulated. The helium diffusivity (DHe) and the ion thermal diffusivity (χi) are 5–15 times higher than the neoclassical level in the high βp mode plasma. In the RS plasma, DHe is reduced from 6–7 times to a 1.4–2 times higher level than the neoclassical level when χi is reduced from 7–18 times to a 1.2–2.6 times higher level than the neoclassical level. The carbon and argon diffusivities estimated assuming the neoclassical inward convection velocity are 4–5 times larger than the neoclassical value, even when χi is close to the neoclassical level. Argon exhaust from the inside of the ITB is demonstrated by applying electron cyclotron heating (ECH) in the high βp mode plasma, where both electron and argon density profiles become flatter. The flattening of the argon density profile is consistent with the reduction of the neoclassical inward convection velocity due to the reduction of the bulk plasma density gradient. In the RS plasma, the density gradient is not decreased by ECH and argon is not exhausted. These results suggest the importance of density gradient control in suppressing impurity accumulation.

A substantial reduction of the L-H transition threshold power was observed under the W-shaped divertor in JT-60U, in comparison with the open divertor. Radiation power in the main region was also decreased in this case. These differences of radiation power from the main plasma between open and W-shaped divertor were not large enough to account for the apparent reduction of threshold power. In this paper, the emphasis is on the edge plasma parameters which relate to the L-H transition. It was found that edge plasma pressure just before the L-H transition in the W-shaped divertor case became smaller than that in the open divertor case. After the modification of divertor geometry, edge ion collisionality just before the L-H transition also became lower and it was established with lower auxiliary power input. It was suggested that reduction of edge ion collisionality arose from neutral particles near the X-point by the analysis of poloidal profile of neutral density. Therefore, neutral particles near the X-point prevented the threshold power for the L-H transition from more reduction at low plasma density, and caused the shift of plasma density which gave minimum threshold power.

Recent lower hybrid current drive (LHCD), and heating (LHH) experiments on JT-60 are reported. The current drive product of neRpIRF approximately 12.5*1019 m-2 MA was achieved at the LH power of approximately 4.5 MW, and the CD efficiency, the energy confinement, the global power balance and the heat load on divertor plates were investigated in high power LHCD plasmas. Nearly steady state H-mode discharges were found during LHCD with two different frequency injections. Sawtooth suppression in NB heated plasmas by LHCD have shown an improvement in confinement near the plasma center. Parametric instabilities in LH heating experiments were significantly reduced by increasing the plasma current, and the stored energy increased linearly with heating power of up to approximately 9 MW at ne approximately 7*1019 m-3 and Ip=2.75 MA. Parametric instabilities near the plasma edge in the ion heating regime were also reduced in peaked density plasmas produced by pellet injection and LH waves increased the central plasma pressure at ne(0) > 1.4*1020 m-3.

Our understanding of the physics of internal transport barriers (ITBs) is being advanced by analysis and comparisons of experimental data from many different tokamaks worldwide. An international database consisting of scalar and two-dimensional profile data for ITB plasmas is being developed to determine the requirements for the formation and sustainment of ITBs and to perform tests of theory-based transport models in an effort to improve the predictive capability of the models. Analysis using the database indicates that: (a) the power required to form ITBs decreases with increased negative magnetic shear of the target plasma, and: (b) the E x B flow shear rate is close to the linear growth rate of the ion temperature gradient (ITG) modes at the time of barrier formation when compared for several fusion devices. Tests of several transport models (JETTO, Weiland model) using the two-dimensional profile data indicate that there is only limited agreement between the model predictions and the experimental results for the range of plasma conditions examined for the different devices (DIII-D, JET, JT-60U). Gyrokinetic stability analysis (using the GKS code) of the ITB discharges from these devices indicates that the ITG/TEM growth rates decrease with increased negative magnetic shear and that the E x B shear rate is comparable to the linear growth rates at the location of the ITB.

A conceptual design of a steady-state tokamak, JT-60 Super Upgrade, is being carried out. The present JT-60 facility will be fully utilized for this upgrade. The major objective of JT-60SU is to establish an integrated basis of physics for steady-state tokamak reactors such as SSTR. JT-60SU aims at investigation of steady-state operation relevant to the reactor regime. The technology development for long pulse operation and research for fusion safety improvement are also aimed at. The toroidal field coils are designed with using an advanced disk structure to improve the force transmission to the structure material. A modified three-in-hand grading using a circular Nb 3 Al and NbTi cable-in-conduit conductors is also adopted. A 40 cm thick water tank-type vacuum vessel made of Ti-6Al-4V, and boron (1% of 10 B) doped water is chosen for the vacuum vessel to reduce activation and to increase neutron shielding. Through these studies, we plan to conduct advanced tokamak experiments in JT-60SU for steady-state tokamak reactors. These studies will also contribute to the steady-state bum in ITER as an ultimate goal.

Emphases in JT-60 experiments are placed on (1) lower-hybrid (LH) current drive characteristics with a multi-junction type launcher, and (2) the confinement study with combination of neutral beam injection LH current drive and pellet injection. The new multi-junction LH launcher provides a sharp N/sub /// spectrum with high directivity for N/sub ///=1-3.4. The current drive efficiency and the radial distribution of high energy electron production show clear correlation with injected N/sub ///: the current drive efficiency has the maximum at low N/sub ///( approximately 1.3) while flattening of plasma current is more effective in large N/sub ///. A broad radial distribution of high energy electron current and approximately 30% reduction in sawtooth inversion radius were obtained by high N/sub /// ( approximately 2.5) LH injection. To fully suppress the sawtooth activity, low N/sub /// ( approximately 1.3) injection was found to be more effective. Improved energy confinement has been obtained with hydrogen pellet injection. Energy confinement time was enhanced up to 40% relative to usual gas fuelled discharges. The discharge has a strongly peaked electron density profile with ne(0)/(ne) approximately 5 and ne(0) approximately 2.0*1020 m-3. The improved discharges are characterized by a strongly peaked pressure profile within the q=1 magnetic surface, and degrades when a large sawtooth recovers or the pressure gradient may reach a critical value. When large (3 mm, 4 mm) and fast (2.2 km/s) pellets were injected, 30% energy confinement improvement was obtained even during the NB heating of 14 MW. Further investigations of IDC characteristics have been made. The oxygen impurity lines from the main plasma and the main radiative loss drop first. Then the plasma stored energy starts to rise. The particle recycling is reduced around the main plasma, and is localized in the neighborhood of the X-point with a time lag of approximately 0.2 sec. Eventually the discharge shows a significant remote radiative cooling power at the divertor region.

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.

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

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.

The possibility of achieving a high effective Q plasma (Q is the fusion power multiplication factor) by increasing the power transfer from alpha particles to the background plasma through heating in the ion cyclotron range of frequencies (ICRF) is examined. The effective Q-value, Qeff, which is defined as the ratio of the indirect plasma heating power from ICRF heated alpha particles to the direct heating power, is evaluated by using the rate of wave absorption by alpha particles on the basis of the energy moments of the quasi-linear RF diffusion operator. The fourth harmonic alpha cyclotron wave, which is absorbed mainly by alpha particles in the absence of hydrogen, should be favourable for achieving a high Qeff plasma. In the high magnetic field, Qeff can also be quite well increased by the third harmonic wave. Effects of tail formation by ICRF waves and of wave absorption by hydrogens on Qeff are also discussed.