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The applicability of tokamak, helical and laser fusion reactors as a volumetric fusion neutron source has been examined using the systems codes that have been utilised for the conceptual design of DEMO and commercial reactors in Japan. This study has clarified the characteristics of reactor-based volumetric neutron sources that can be designed based on the current physics and engineering basis with a reasonable running cost (∼5B Yen/year). Although the achievable neutron flux is 2–3 orders lower than that of accelerator driven neutron sources, tokamak and helical neutron sources can provide a much larger irradiation area for the test of large components. Laser neutron sources have both high operability and tritium breeding capacity. These reactor-based neutron sources also serve as an integrated test bed of the entire reactor system.

The operation scenarios for plasma breakdown and current ramp-up phases in JT-60SA tokamak have been developed and verified in simulation using the TOSCA code. The induced current in the conducting elements such as vacuum vessel and stabilizing plate increases to the comparable level of plasma current of ∼600 kA during the breakdown phase and thus enhances the strength of stray field. The optimized scenarios for half and full pre-magnetization cases satisfied the conditions required for the plasma initiation. At the initial plasma, the vertical magnetic field required to sustain the plasma position was controlled by the outer equilibrium field (EF) coil currents which compensate for a vertical field due to a large eddy current. The condition for the formation of divertor configurations given by the combination of the magnetic flux for plasma and the plasma current has been satisfied which enables us to develop the operational scenarios with a smooth transition from a limiter to a divertor configuration.

The differences and similarities between the internal transport barriers (ITBs) of tokamak and helical plasmas are reviewed. By comparing the characteristics of the ITBs in tokamak and helical plasmas, the mechanisms of the physics for the formation and dynamics of the ITB are clarified. The ITB is defined as the appearance of discontinuity of temperature, flow velocity, or density gradient in the radius. From the radial profiles of temperature, flow velocity, and density the ITB is characterized by the three parameters of normalized temperature gradient, $R/{L}_{T}$, the location, ${\rho }_{\mathrm{ITB}}$, and the width, W/a, and can be expressed by 'weak' ITB (small $R/{L}_{T}$) or 'strong' (large $R/{L}_{T}$), 'small' ITB (small ${\rho }_{\mathrm{ITB}}$) or 'large' ITB (large ${\rho }_{\mathrm{ITB}}$), and 'narrow' (small W/a) or 'wide' (large W/a). Three key physics elements for the ITB formation, radial electric field shear, magnetic shear, and rational surface (and/or magnetic island) are described. The characteristics of electron and ion heat transport and electron and impurity transport are reviewed. There are significant differences in ion heat transport and electron heat transport. The dynamics of ITB formation and termination is also discussed. The emergence of the location of the ITB is sometimes far inside the ITB foot in the steady-state phase and the ITB region shows radial propagation during the formation of the ITB. The non-diffusive terms in momentum transport and impurity transport become more dominant in the plasma with the ITB. The reversal of the sign of non-diffusive terms in momentum transport and impurity transport associated with the formation of the ITB reported in helical plasma is described. Non-local transport plays an important role in determining the radial profile of temperature and density. The spontaneous change in temperature curvature (second radial derivative of temperature) in the ITB region is described. In addition, the key parameters of the control of the ITB and future prospects are discussed.

The modification of JT-60 to a fully superconducting coil tokamak, JT-60SA (JT-60 Super Advanced) device, has been programmed to contribute and supplement ITER toward to DEMO. Lower divertor design with the ITER-like lower single null divertor configuration is studied to obtain high heat and particle controllability using SOLDOR/NEUT2D code. With anticipated total power flux into SOL of 37 MW (90% of input power), the peak heat load on outer divertor target can be reduced to 5.8 MW/m 2 at the detached condition by gas puffing in the vertical divertor target with the “V-shaped corner”. It is ∼2/5 of the allowable level of 15 MW/m 2 . On the other hand, particle controllability such as control of detached to attached condition by divertor pumping is improved by increase the strike point distance from 20 to 120 mm with above divertor geometry, suggesting that recover from severe detachment at the small distance case can be achieving by elevation of the strike point locations. Optimization of upper divertor design is in progress for high β steady-state operation using upper single null divertor configuration.

The application of the electron cyclotron heating (ECH) and electron cyclotron current drive (ECCD) to the JT-60U tokamak started in 1999. Because the power deposition by the electron cyclotron wave is very localized and controllable, the application ofECH/ECCD has been very attractive for the following recent studies in the JT-60U: (a) the extension of plasma performance toward high normalized beta (β N ), (b) high bootstrap current fraction, and (c) long-pulse operation (65 s). Plasma produced in the studies aiming at advanced steady tokamak is considered to be in a kind of "self-organized state" with external input power by joule heating plus additional heating. The internal transport barrier that develops by the additional heating enhances the local bootstrap current by steep pressure gradient, and then the modified plasma current profile establishes a different confining poloidal magnetic field configuration from the initial configuration. In such experimental research in the JT-60U, the ECH contributes as an active tool for the plasma control to study the physical mechanisms of high-IS magnetohydrodynamic instability, internal transport barrier, current hole, and so on. Results of the ECH/ ECCD applications in the JT-60U are briefly reviewed.

Predictions of the recently developed paleoclassical transport model are compared with data from many toroidal plasma experiments: electron heat diffusivity in DIII-D, C-Mod and NSTX ohmic and near-ohmic plasmas; transport modeling of DIII-D ohmic-level discharges and of the RTP ECH 'stair-step' experiments with eITBs at low order rational surfaces; investigation of a strong eITB in JT-60U; H-mode Te edge pedestal properties in DIII-D; and electron heat diffusivities in non-tokamak experiments (NSTX/ST, MST/RFP, SSPX/spheromak). The radial electron heat transport predicted by the paleoclassical model is found to agree with a wide variety of ohmic-level experimental results and to set the lower limit (within a factor {approx} 2) for the radial electron heat transport in most resistive, current-carrying toroidal plasmas -- unless it is exceeded by fluctuation-induced transport, which often occurs in the edge of L-mode plasmas and when the electron temperature is high ({approx}>T{sub e}{sup crit} {approx}B{sup 2/3}{bar {alpha}}{sup 1/2} keV) because then paleoclassical transport becomes less than gyro-Bohm-level anomalous transport.

High densities exceeding the Greenwald limit by a factor of 1.7 have been obtained in L-mode discharges with high internal inductances of elli as high as 2.8 in JT-60U. The internal inductance is controlled by ramping down the plasma current. In addition to the extension of the operational regime limited by disruptions, confinement performance remains as good as an H89PL factor of 1.6 even above the Greenwald limit. While an earlier high elli study has indicated core confinement improvement due to enhancement of the poloidal field, the additional improvement of the tolerance against the high density turned out to be correlated with high edge temperature. The normalized density when the detachment occurs, characterized by a decrease in the Dα signal at the divertor, is even higher in the case with no disruption than in the case with a disruption. These comparisons have indicated that the improvement in thermal and particle transport does exist in the periphery and in the edge in high elli plasmas, and the shift of the density limit towards higher densities is observed coincidently. Although the high elli discharge studied here lies outside the usual parameter space for steady-state operation of a tokamak, demonstration of a stable discharge with good confinement beyond the Greenwald limit suggests that the magnetic shear at the edge is one key parameter to uncover the physical elements of the operational density limit.

Long-spatial-scale density fluctuations with a wave number of the order of 1 cm−1 were suppressed in the internal transport barrier (ITB) region of a JT-60U reversed magnetic shear plasma, after a cold pulse was induced by the shallow pellet injection at the plasma edge. The particle and ion heat diffusivities are reduced with the suppressed fluctuations. The reduction of the electron heat diffusivity is observed only transiently in the inner ITB portion around the layer of the density fluctuation measurement during the cold pulse propagation and it is slower (~ 15 ms) than the reduction of the density fluctuation level (< 5 ms). In the outer ITB portion, the electron heat diffusivity increases transiently. These results indicate decoupling of the electron heat transport and the long-spatial-scale density fluctuations.

Further transport reduction is induced by a pellet or electron cyclotron (EC) wave injection after the formation of an internal transport barrier in JT-60U reversed shear plasmas. After pellet injection, the O-mode reflectometer signal drastically decreases, indicating reduction in the electron density fluctuation level. The ion thermal diffusivity decreases by one order of magnitude and reaches the neoclassical level. The effective particle diffusivity also decreases by one order of magnitude, but the electron thermal diffusivity remains constant. After EC wave injection, when no reduction of the O-mode reflectometer signal is observed, the electron thermal diffusivity decreases by a factor of ~2–3, while decreases in the ion thermal diffusivity and the effective particle diffusivity are small. These results indicate that the ion heat transport and particle transport were coupled with the measured density fluctuation, but electron heat transport was decoupled. Ion heat and particle transport and electron heat transport were dominated by different types of fluctuation.

Characteristics of the edge pedestal has been studied for the high βp ELMy H mode and the reversed shear ELMy H mode in JT-60U. The contribution of the pedestal is around 20% for the confinement improvement factor, the normalized beta, and the bootstrap fraction. The upper boundary of the fusion performance measure increases with the pedestal poloidal beta, βp-ped. At high triangularity, δ, βp-ped increases with the total βp (βp − tot) almost linearly for positive shear type I ELMy discharges. This dependence is not due to the profile stiffness, since the dependence is the same for the discharges both with and without an internal transport barrier (ITB). In the reversed shear ELMy H mode, βp-ped increases with βp-tot on the same line as the positive shear cases except at high q95 = 8–9.3. In the high plasma current regime, the final structure with both an ITB and edge transport barrier seems to be determined by the balance between the expanding ITB-foot radius and deepening ELM penetration: the ELM penetration radius deepens with increasing pedestal stored energy and then reaches the ITB-foot radius. The ITB-foot seems to behave as a barrier against the ELM crash penetration, and shrinks after a few ELM attacks, and the ELM penetration follows the shrinking ITB-foot.

The Large Helical Device is a heliotron device with L = 2 and M = 10 continuous helical coils with a major radius of 3.5–4.1 m, a minor radius of 0.6 m and a toroidal field of 0.5–3 T, which is a candidate among toroidal magnetic confinement systems for a steady state thermonuclear fusion reactor. There has been significant progress in extending the plasma operational regime in various plasma parameters by neutral beam injection with a power of 13 MW and electron cyclotron heating (ECH) with a power of 2 MW. The electron and ion temperatures have reached up to 10 keV in the collisionless regime, and the maximum electron density, the volume averaged beta value and stored energy are 2.4 × 1020 m−3, 4.1% and 1.3 MJ, respectively. In the last two years, intensive studies of the magnetohydrodynamics stability providing access to the high beta regime and of healing of the magnetic island in comparison with the neoclassical tearing mode in tokamaks have been conducted. Local island divertor experiments have also been performed to control the edge plasma aimed at confinement improvement. As for transport study, transient transport analysis was executed for a plasma with an internal transport barrier and a magnetic island. The high ion temperature plasma was obtained by adding impurities to the plasma to keep the power deposition to the ions reasonably high even at a very low density. By injecting 72 kW of ECH power, the plasma was sustained for 756 s without serious problems of impurities or recycling.

The real-time feedback control of the plasma stored energy has been developed for control of the plasma MHD stability in the JT-60U tokamak. The plasma stored energy can be detected with high accuracy in real-time by a function parameterization method for various plasmas available in JT-60U, such as Ohmic plasmas, the L-mode, the H-mode, the high poloidal beta mode and the reversed shear mode over a wide range of the plasma parameters. By manipulating the neutral beam injection power, the plasma stored energy has been successfully controlled along the preprogrammed reference waveform. Especially in the reversed shear mode, this feedback control scheme has improved the reproducibility of the formation of the internal transport barrier, and MHD activities could be suppressed keeping the normalized beta in a stable region. A DT equivalent fusion amplification gain of 0.5 was sustained for 0.8 s in a reversed shear plasma by employing this feedback control scheme.

The diagnostic system of JT-60U (JT-60upgrade) is composed of about 50 individual diagnostic devices. Recently, the detailed radial profile measurements of plasma parameters have been improved, so ...

The significance of real-time feedback control is emphasized in this paper as an indispensable method to improve and sustain the improved plasma characteristics in JT-60U high fusion performance discharges as well as to operate the fusion reactor under the optimal divertor conditions with respect to the heat load and exhaust pumping. In accordance, substantial improvement in the equivalent fusion amplification gain of over unity has been reproducibly achieved at the JT-60U tokamak in the reversed shear mode of operation with the robust feedback controls, where the value of target density was deliberately optimized for the reliable internal transport barrier formation, and the magneto-hydrodynamic stability control was performed with the stored energy feedback. The feedback control techniques also demonstrated the effectiveness to produce quasi-steady-state high-performance plasmas. In addition, three major parameters associated with the fusion reactor instrumentations, namely the neutron production rate, operating density, and divertor radiation power, were simultaneously feedback controlled in the ELMy H-mode plasmas. Here, the matrix response function was evaluated to identify the limitations involved with the linear combination of independent controls. Other advanced feedback schemes, such as the feedback suppression of the neoclassical tearing mode required to sustain high plasma pressure in a steady-state, are alsomore » described. Finally, the controversial issues for the future intelligent plasma control necessary for the advanced steady-stated tokamak reactor are addressed.« less

Fusion plasma performance and confinement studies on JT-60 and JT-60U are reviewed. With the main aim of providing a physics basis for ITER and the steady-state tokamak reactors, JT-60/JT-60U has b...

Non-inductive full current drive (CD) in a reversed magnetic shear (RS) plasma with high energy confinement of 1.4 times the H-mode scaling at 82% of the Greenwald density has been achieved simultaneously for the first time in a tokamak; these results were obtained in the JT-60U tokamak. The full-CD state was realized by a combination of waves in the lower hybrid range of frequency and neutral beam injection together with high bootstrap current fraction (fbs) of 62% to the total plasma current. The combination has demonstrated for the first time active current profile modification in a high fbs RS plasma both in the core and in the peripheral regions.

Although both the transport barriers in the plasma interior and edge, namely internal transport barrier (ITB) and H-mode, are simultaneously required to sustain high-β and improved confinement in steady state, experimental observations suggest that theL-H transition and ITB formation are independent phenomena with different formation conditions, and they often exhibit apparent contradictions to each other. Accordingly, the compatibility conditions of edge and internal barrier formation have been investigated in JT-60U. The L-H threshold power is substantially increased in strong ITB plasmas, and high-performance discharges often persist in L-mode. On the other hand, the ITB formation power increases in H-mode plasmas. It was found, however, that the L-H transition criteria expressed in terms of the edge quantities are similar in plasmas with and without the ITB, and the edge density is extremely low in strong ITB plasmas. It has also been demonstrated in JT-60U that the controlled degradation of the ITB quality, in terms of the changes of toroidal torque input, results in an increase of edge density and induces theL-H transition. Therefore, the quality of ITB modifies the L-H transition threshold power, indicating that the additional `hidden' parameter is present in ITB plasmas. Other approaches of dynamically controlling the L-H transition in ITB plasmas, such as the triangularity ramp developed in JT-60U, are also discussed in this paper.

Toward the study of steady-state operation of fusion plasma, it is important to assess the confinement characteristics under a condition in which fractional power of electron heating is dominant to the total heating power. In a fusion plasma electron heating by α particles is dominant. Especially, the influence on the internal transport barrier (ITB) is an important issue, since formation of ITB is a key issue in advanced steady-state operation from a viewpoint of both attaining highly improved confinement and larger bootstrap current fraction to the total current. In JT-60U, the influence on ITBs when fractional power of electron heating to the total power is increased has been investigated by using negative-ion based neutral beam injection and electron cyclotron range of frequency (ECRF). It is found that in a reversed magnetic shear plasma of high confinement improvement, still higher over all confinement can be achieved at electron heating powers of 50-60% to the total absorbed power. Reduction of the ion temperature at ECRF injection is observed in an ITB plasma of positive shear. And with intense electron central heating on lower hybrid current drive plasma, high central electron temperature with ITB has been formed.

The dominant causes of the reduction in the enhancement factor of thermal energy confinement with increasing plasma density in ELMy H mode plasmas in JT-60U were analysed by separating the thermal stored energy into core and pedestal components. The enhancement factor of the core confinement evaluated by the offset non-linear scaling which was developed in the ITER H mode confinement database decreased significantly with increasing density, while the pedestal component remained almost constant over a wide range of density. The influence of the edge pedestal structure imposed by the destabilization of type I ELMs on thermal energy confinement of the core plasma was also quantitatively clarified. As the density increased, the saturation of the pressure at the pedestal shoulder, pped, caused by the type I ELM activities forced a reduction in the pedestal temperature. The core temperature for each species, in turn, decreased only by an approximately constant factor with a reduction in the pedestal temperature, resulting in the saturation of the thermal stored energy in the high density regime.

The formation properties of an internal transport barrier (ITB) were investigated in a weak positive magnetic shear plasma by changing the neutral beam heating power. The ion thermal diffusivity in the core region shows L-mode state, weak ITB, and strong ITB, depending upon the heating power. Two features of ITB formation were experimentally confirmed. Weak ITB was formed in spite of the absence of an apparent transition in an ion temperature profile. On the other hand, strong ITB appeared after an apparent transition from the weak ITB. In addition, the ion thermal diffusivity at the ITB is correlated to the radial electric field shear. In the case of the weak ITB, ion thermal diffusivity decreased gradually with increases in the radial electric field shear. There exists a threshold in the radial electric field shear, which allows for a change in state from that of weak to strong ITBs.

Nationally coordinated research on spherical tokamak is being conducted in Japan. Recent achievements include: (i) plasma current start-up and ramp-up without the use of the central solenoid by RF waves (in electron cyclotron and lower hybrid frequency ranges), (ii) plasma current start-up by AC Ohmic operation and by coaxial helicity injection, (iii) development of an advanced fuelling technique by compact toroid injection, (iv) ultra-long-pulse operation and particle control using a high temperature metal wall, (v) access to the ultra-high-β regime by high-power reconnection heating, and (vi) improvement of spherical tokamak plasma stability by externally applied helical field.

Evolution of the ion temperature gradient parameter (ηi) has been studied in JT-60U reverse shear (RS) discharges experimentally. The ηi(r,t) behaviours for RS with and without internal transport barrier (ITB) discharges were observed, where the ITB appeared as steep gradients in the ion temperature profile, the electron density and temperature profiles. The relationship between the electron density scale length (Lne) and the ion temperature scale length (LTi) for different heating methods (NBI only and NB{} + {}ICRF) in RS discharges was compared with each other. It was observed that ηi is a key parameter to characterize ITB formation for the two cases. Kinetic stability for the toroidal drift-type modes (s

The current drive capability of negative ion based neutral beam injection (N-NBI) in JT-60U has been extended to the reactor relevant regime. The driven current profile and current drive efficiency have been evaluated in a high electron temperature regime Te(0) ≈ 10 keV, and reasonable agreement with the theoretical prediction has been confirmed in this regime. The N-NB driven current reached 1 MA with an injection power of 3.75 MW at a beam energy of 360 keV. A current drive efficiency of 1.55 × 1019A m-2 W-1, approaching the ITER requirement, was achieved in the high βp H mode plasma with Te(0) ≈ 13 keV. This current drive performance permitted sustainment of a high beta (βN = 2.5) and high confinement (HHy2 = 1.4) plasma in the full current driven condition at a plasma current of 1.5 MA. The influence of instabilities on the N-NBI current drive was studied. When a burst-like instability driven by N-NBI occurred in the central region, reductions in loop voltage near the magnetic axis and in the neutron production rate due to loss of beam ions were observed although the lost driven current was at most ~7% of the total driven current. When a neoclassical tearing instability appeared in high beta plasmas, the loss of beam ions was enhanced with increasing instability activity.

Internal transport barrier (ITB) formation in the positive shear zone of reverse shear (RS) plasmas in the JT-60U tokamak is described as a series of ITB events, which are abrupt in time and wide in space variations of heat diffusivity, observed as the simultaneous rise and decay of the electron temperature Te in two zones. A new source of heat pulse propagation (HPP) is found in RS plasmas. HPP is created by an ITB event. For the last ITB event described in the present paper, the region of strong (~20 keV s-1) Te rise is initially well localized (~4 cm) in space. Later, a slow diffusive broadening of the rising Te perturbation is seen. Outward HPP is analysed in the region with ~8 cm width fully located in the positive shear space zone. Values of the electron dynamic heat diffusivity as low as ~0.1 m2 s-1 are found. A similar low value of the ion dynamic heat diffusivity (close to the neoclassical value) is obtained for ion HPP. An important consequence of HPP analysis is the absence of electron and ion `heat pinch' in the ITB region.

The effects of radial electric field and plasma rotation on the characteristics of internal transport barriers (ITBs) and core confinement properties are studied in JT-60U reversed shear plasmas. The combination of on-axis and off-axis tangential NBI in the co- and counter-directions allows various profiles of the toroidal rotation. With the same core absorbed power, the case of a small momentum source at the ITB layer is more favourable for sustainment of the ITB layer and the retention of the core energy confinement than the cases of the excess co- or counter-momentum source.

The current drive and heating properties of negative ion based NBI have been studied comprehensively in JT-60U. It has been confirmed from shine-through measurements of the injected beam (350 keV) that multistep ionization processes are essential in the ionization processes of high energy particles. The profile of the current density driven by a negative ion based NB (N-NB) has been determined experimentally. This is in good agreement with the theoretical prediction, and N-NB driven current reached 0.6 MA with EB = 360 keV and PINJ = 3.7 MW. The current drive efficiency ηCD is increased by increasing electron temperature and improved by increasing beam energy. The fast ions from N-NBs are well confined in the enhanced confinement core by the weak poloidal magnetic field of reversed shear plasmas. A clear H mode transition was obtained with N-NB dominant heating, where the net absorbed power required for an H mode transition seemed similar to the previous result obtained in JT-60U using a low energy beam (90 keV). With the strong electron heating by N-NBI (80% absorbed by electrons), an H factor ( = τE/τITER-89PLE) of 1.64 with Te(0) = 1.4Ti(0) was obtained in the steady state ELMy phase.

New features of the space-time evolution of the internal transport barrier (ITB) were highlighted during recent JT-60U reverse shear (RS) experiments. An ITB evolution in RS plasmas is often a combination of the fast time-scale processes and the gradual ones. Fast time-scale processes are the common intrinsic features of JT-60U RS plasmas and are seen as the simultaneous (within a few milliseconds) rise and decay of electron temperature (Te) on two zones separated by a region without variation of Te. The region without variation of Te is located near the position of the minimum safety factor profile (qmin) for many fast processes. The present robust result is that the region of fast-time-scale improvements of electron heat diffusivity is wide in space (around 0.3 of the minor radius) and well extended to the zone of Te decay.

Plasmas with an internal transport barrier (ITB) were heated with a second-harmonic ion cyclotron range of frequency (ICRF) minority heating during the reversed magnetic shear (RS) experiments in JT-60U. In order to control the central particle fuelling, the input power of the neutral beam injection (NBI) heating during ICRF heating was scanned shot-by-shot. The electron temperature gradient around the ITB increased with the total input power. Steeper gradients were obtained with additional ICRF heating by avoiding collapses induced by the steep pressure gradient. The difference of the time derivative at the ICRF injection timing indicated a fast ion loss of about 40% during the RS operation. The density gradient was increased with the central fuelling by the NBI heating and the shoulder position of the slope at the ITB moved outward with the particle fuelling. The particle confinement time at the ITB increased with the central particle fuelling while the particle confinement time of the low-fuelling discharges were similar to that of the normal shear L-mode discharge.

The particle transport for several impurity ion species was investigated in reversed shear and ELMy H-mode plasmas on JT-60U. Particle diffusivity for neon was evaluated from gas-puffing modulation experiments; the diffusivity was smaller by one order of magnitude at the internal transport barrier in reversed shear plasmas than in ELMy H-mode plasmas. An increase of the inward pinch velocity for neon was observed at the internal transport barrier in the reversed shear plasma. The particle transport coefficients at the internal transport barrier were compared for neon, carbon and helium. The particle diffusivity was almost the same for all ions, while the inward pinch velocity was larger for ions having higher charge state.

Characteristics of energetic ion tail formation during the second-harmonic ion cyclotron range of frequency (2CH ICRF) heating and confinement of tail ions were investigated in negative magnetic shear plasmas on JT-60U. Charge-exchange neutral particle measurements show significantly lower tail temperature in negative-shear plasmas than those in positive-shear plasmas. The tail temperature in negative-shear plasmas increases with plasma current and shows improvement of energetic ion confinement. Charge-exchange neutral fluxes are also lower in negative-shear plasmas than in positive-shear plasmas. Meanwhile, tail ion stored energy is comparable between negative- and positive-shear plasmas. In order to understand these experimental results, simple modelled calculations were carried out with an orbit-following Monte Carlo (OFMC) code. The model calculations suggest that the ICRF power absorption is dominant near the 2CH resonance in the negative-shear plasma and in the low-field side away from the 2CH resonance in the positive-shear plasma. A larger scale length of pressure profile in the negative-shear plasma compensates for a lower density of tail ions than that in the positive-shear plasma. Accordingly, negative-shear plasmas can contain comparable stored energy of energetic tail ions to positive-shear plasmas.

Reduction of electron thermal diffusivity was investigated by comparing the internal transport barrier (ITB) with that of high- discharges. In reversed shear discharges heated by neutral beams, reduction was significant when negative magnetic shear was strong. Radial location of ITB propagated outwards and then localized near the region of q-minimum for both the ion and the electron channels. Reduction in was also observed in the electron heating regime by ion cyclotron heating

The particle diffusivity and the convection velocity in the reversed shear plasma of JT-60U were separately evaluated based on the perturbation technique using modulated helium gas-puffing. The particle diffusivity in the region of internal transport barrier was reduced by about a factor of two compared with that in the inside region. The inward convection velocity was observed in the region of internal transport barrier, while the outward convection velocity was observed in the inside region. These results indicate that the changes in both the particle diffusivity and the convection velocity are related to the formation of the internal transport barrier.

Negative central shear discharges in JT-60U revealed an internal transport barrier (ITB) which is located at larger minor radii than those in other devices and is accompanied by a clear reduction in . In order to investigate this origin of the unique and favourable feature for future reactors, the discharge parameters were compared with those in DIII-D, where ITB was at smaller minor radii and less reduction in was observed. Strong negative magnetic shear was found to be related to the reduction in . The thin skin depth in JT-60U originating from high , together with the continuous decrease in , was effective in moving the ITB location to larger minor radii and in sustaining the strong negative shear.

Isolation of the core with improved confinement (inside the internal transport barrier) from high recycling and strongly radiative boundary plasmas was demonstrated for the first time in the reversed shear plasmas of the JT-60U tokamak. The internal transport barrier remained intact, while edge radiation and recycling was enhanced with neon and hydrogen gas puffing. The internal transport barrier reduced the particle transport by a factor 10{endash}15, while shrinking in the time scale of resistive diffusion towards the center. {copyright} {ital 1997} {ital The American Physical Society}

In the JT-60U tokamak a pre-formed reversed magnetic shear configuration was sustained non-inductively by means of lower hybrid current drive (LHCD). It was formed by neutral beam heating at current build-up and had a reversed shear region extending to 65% of the minor radius. The configuration was sustained for 7.5 s, and the extent of the reversed shear region was kept as large as 55% of the minor radius until the very end. The plasma was stable against MHD activities. It is also demonstrated that LHCD could modify reversed magnetic shear keeping plasma parameters, such as the plasma current and the toroidal magnetic field, the same.

By modification of current and heat deposition profiles, the integrated core plasma performance was improved in ELMy H-mode. In a quasi steady state, beta p approximately 2.5-3, beta N approximately 2.5-3.1 and H-factor approximately 1.8-2.1 were sustained simultaneously under the full current drive condition (bootstrap current 55%, beam driven current 45% at Ip=0.5MA). The achievable beta p value and the energy confinement performance increases systematically with safety factor at the edge qeff and a high qeff operation (qeff>6) is required to obtain favorable integrated performance. To obtain these improved high beta ELMy plasmas, it is essential to control MHD activities; suppression of beta p-collapse and control of ELM activity.

Spontaneous formation of internal transport barrier was observed in JT-60U high- beta p discharges, where q=3 surface is the most likely explanation for its radial location. Further improved confinement was brought about by the subsequent transition to the edge transport barrier.

Improvement of an enhanced confinement state in a high poloidal beta (βp) regime without sawtooth activity has been achieved in JT-60U. A confinement mode has been demonstrated where both the edge and the core confinement are improved. The attainable βp was also extended to higher values in this improved mode, because of its broader pressure profile. As a result of the improvement in confinement and in attainable βp, the highest value of the fusion triple product has been extended by a factor of 2.5 over that achieved in the 1992 experiments; it has reached (1.1 ± 0.3) × 1021 m-3.s.keV with a central ion temperature of about 37 keV. The D-D neutron emission rate has also been doubled in these experiments and has reached (5.6 ± 0.6) × 1016 s-1

Toroidal field ripple loss of neutral-beam-injected fast ions is measured from the heat load on the first wall in JT-60U. The heat load is localized in both toroidal and poloidal directions and increases with both ripple size and safety factor. The overall measured ripple loss is in good agreement with an orbit-following Monte Carlo calculation. Nevertheless, there is a small difference between the experimental and the calculated location of the maximum heat load, which suggests that the radial electric field shifts the impinging points of ripple-trapped ions on the first wall.

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.

The design of the modification of JT-60U, JT-60SA has been optimized from the viewpoint of plasma performance, and operation regimes have been evaluated with the latest design. Upper and lower divertors with different geometries will be prepared for flexibility of the plasma shape, which will enable both low aspect ratio (A ~ 2.65) and ITER shape (A = 3.1) configurations. The beam lines of negative-ion neutral beam injection will be shifted downwards by ~0.6 m for the off-axis current drive (CD), in order to obtain a weak/reversed shear plasma, as well as having the capability of heating the central region. The feedback control coils along the openings in the stabilizing plate are found effective in suppressing the resistive wall mode and sustaining high βN close to the ideal wall limit. Sustainment of plasma current of 3–3.5 MA for 100 s will be possible in ELMy H-mode plasmas with moderate heating power, βN, and density within an available flux swing. It is also expected that higher βN, high-density ELMy H-mode plasmas will be maintained for 100 s with higher heating power. The expected regime of full CD operation has been extended with upgraded heating and CD power. Full CD operation for 100 s with reactor-relevant high values of normalized beta and bootstrap current fraction (Ip = 2.4 MA, βN = 4.3, fBS = 0.69, , HH98y2 = 1.3) is expected in a highly-shaped low-aspect-ratio configuration (A = 2.65).