Table of contents

Stabilization of an m = 3/n = 2 neoclassical tearing mode (NTM) has been studied experimentally by applying a local electron cyclotron current drive (ECCD) in the JT-60U tokamak. The EC power is injected before the mode onset, and its effect is compared with the ECCD applied at the saturation phase. The experimental results show that the ECCD applied at the growth phase is more effective than that applied at the saturation phase. The necessary EC power for the suppression is reduced and the mode onset is delayed, indicating the hysterisis characteristics of the NTM on the ECCD stabilization. The dependence on the EC power and injection angle is also shown.

Investigations of the radial electric field have been carried out at the stellarator Wendelstein 7-AS using charge-exchange spectroscopy with high spatial resolution based on the high-energy Li beam diagnostic (Li-CXS). To evaluate the electric field, the radial force balance equation is used together with the measured profiles of poloidal velocity, density and the temperature of carbon impurities. Results of experiments are compared with predictions of the neoclassical theory. We also use a simple analytic approximation, in which the radial electric field is proportional to the pressure gradient of H⁺ ions of the bulk plasma. This approximation agrees reasonably well with both the measurements and the accurate neoclassical calculations and, therefore, can be used for a fast estimation of the radial electric field.

Two-dimensional fluid simulations of scrape-off layer (SOL) turbulence with non-constrained energy content (flux driven) are characterized by profile relaxation and strong outward bursts of density. The ballistic propagation extends well beyond the e-folding length of the SOL with a Mach number of M_⊥ ∼ 0.04. Turbulence stabilization is achieved by biasing part of the limiter surface. The critical radial extent to achieve this stabilization is derived. This effect governs the size of the biased ring required to insulate the wall from long range bursts of matter. The same characteristic scale also governs the critical size of Langmuir probe tips. For probe tips in excess of this size, the flux tube to the probe is found to be decoupled from the background plasma.

Key DIII-D advanced tokamak (AT) experimental and modelling results are applied to examine the physics and control issues for ITER to operate in a negative central shear (NCS) AT scenario. The effects of a finite edge pressure pedestal and current density are included based on the DIII-D experimental profiles. Ideal and resistive stability analyses demonstrate that feedback control of resistive wall modes by rotational drive or flux conserving intelligent coils is crucial for these AT configurations to operate at attractive β_N values in the range 3.0–3.5. Vertical stability and halo current analyses show that reliable disruption mitigation is essential and mitigation control using an impurity gas can significantly reduce the local mechanical stress to an acceptable level. Core transport and turbulence analyses indicate that control of the rotational shear profile is essential to reduce the pedestal temperature required for high β. Consideration of edge stability and core transport suggests that a sufficiently wide pedestal is necessary for the projected fusion performance. Heat flux analyses indicate that, with core-only radiation enhancement, the outboard peak divertor heat load is near the design limit of 10 MW m⁻². Detached operation may be necessary to reduce the heat flux to a more manageable level. Evaluation of the ITER pulse length using a local step response approach indicates that the 3000 s ITER long-pulse scenario is probably both necessary and sufficient for demonstration of local current profile control.

A sandpile model describing some of the features of plasma turbulent transport dynamics in the L-mode is extended, by adding appropriate new dynamics, to exhibit a transition to enhanced confinement modes. As a result, H-modes with and without edge localized modes (ELMs) can both be obtained by varying the appropriate parameters. Each exhibits features reminiscent of what is observed in confined plasmas. The interplay between an avalanche and a diffusive transport mechanism is shown to be essential, in this context, for the system to display periodic edge ELMing.

For the (non-axisymmetric) stellarator class of plasma confinement devices to be feasible candidates for fusion power stations it is essential that, to a good approximation, the magnetic field lines lie on nested flux surfaces; however, the inherent lack of a continuous symmetry implies that magnetic islands responsible for breaking the smooth topology of the flux surfaces are guaranteed to exist. Thus, the suppression of magnetic islands is a critical issue for stellarator design, particularly for small aspect ratio devices. Pfirsch–Schlüter currents, diamagnetic currents and resonant coil fields contribute to the formation of magnetic islands, and the challenge is to design the plasma and coils such that these effects cancel.

Magnetic islands in free-boundary high-pressure full-current stellarator magnetohydrodynamic equilibria are suppressed using a procedure based on the Princeton Iterative Equilibrium Solver (Reiman and Greenside 1986 Comput. Phys. Commun.43 157) which iterates the equilibrium equations to obtain the plasma equilibrium. At each iteration, changes to a Fourier representation of the coil geometry are made to cancel resonant fields produced by the plasma. The changes are constrained to preserve certain measures of engineering acceptability and to preserve the stability of ideal kink modes. As the iterations continue, the coil geometry and the plasma simultaneously converge to an equilibrium in which the island content is negligible, the plasma is stable to ideal kink modes, and the coils satisfy engineering constraints. The method is applied to a candidate plasma and coil design for the National Compact Stellarator eXperiment (Reiman et al2001 Phys. Plasma8 2083).

This paper reports the first results on the measurement of the radial profiles of plasma poloidal and toroidal rotation performed on the TCABR tokamak, in the collisional regime (Pfirsch–Schluter), using Doppler shift of carbon spectral lines, measured with a high precision optical spectrometer. The results for poloidal rotation show a maximum velocity of (4.5 ± 1.0) × 10³ m s⁻¹ at , (a—limiter radius), in the direction of the diamagnetic electron drift. Within the error limits, reasonable agreement is obtained with calculations using the neoclassical theory for a collisional plasma, except near the plasma edge, as expected. For toroidal rotation, the radial profile shows that the velocity decreases from a counter-current value of (20 ± 1) × 10³ m s⁻¹, at the plasma core, to a co-current value of (2.0 ± 0.9) × 10³ m s⁻¹ near the limiter. An agreement within a factor 2, for the plasma core rotation, is obtained with calculations using the model proposed by Kim, Diamond and Groebner (1991 Phys. Fluids B 3 2050).

In this paper, we analyse the main features of the pulsed poloidal current drive (PPCD) technique, used in the reversed field pinch configuration to achieve improved confinement conditions. In the RFX experiment, PPCD corresponds to a decrease of the magnetic fluctuations, to a peaking of the temperature profile, and to a reduced transport and plasma–wall interaction. A three-dimensional MHD nonlinear code and one-dimensional time-dependent transport models have been applied to study the effect of PPCD on the magnetic and plasma profiles. The three-dimensional MHD simulations show that the external inductive drive pinches and peaks the current profile driving the configuration through a transient phase, where the spontaneous turbulent dynamo action is quenched. The one-dimensional transport codes indicate that the experimental profile modifications associated with PPCD are consistent with a reduction of the stochastic transport.

An approximation to a quasi-omnigenous structure of the magnetic field strength has been numerically obtained by optimization of the magnetic configuration of a conventional heliotron/torsatron. This configuration shows very good collisionless confinement of the guiding centre orbits of fusion α-particles for values of ⟨β⟩ up to 0.05.

ITER operational scenarios with high field side pellet fuelling are considered. The possibility of reducing the energy losses per edge localized mode (ELM) to an acceptable level is discussed. Requirements of the pellet-fuelling system for desirable ELM energy reduction are obtained. Self-consistent transport simulations of pellet-fuelled scenarios reveal the possibility of operation with moderate ELM losses, plasma density below the Greenwald density, high energy multiplication factor Q ∼ 20 and power across the separatrix above the estimated L–H mode power threshold.

The effect of a toroidal current hole on the first orbit (FO) loss and on the collisional loss of alpha particles in JET is investigated. Numerical results of predictive three-dimensional Fokker–Planck modelling of the distribution function of D–T fusion alphas in hollow current JET discharges are presented. If the current hole region is kept reasonably small, it induces only a moderate increase of FO losses as well as of the collisional loss of fast alphas. The current hole effect is shown to be qualitatively equivalent to a reduction of the total plasma current I. Hence, the alpha confinement degradation by the current hole profiles can be compensated by enlarging I.

Rotating magnetic fields (RMFs) have been used to both form and sustain low density, prolate FRCs in the translation confinement and sustainment (TCS) facility. The two most important factors governing performance are the plasma resistivity, which sets the maximum density for which toroidal current can be maintained, and the energy loss rate, which sets the plasma temperature. The plasma resistivity has been determined by carefully measuring the amount of RMF power absorbed by the FRC. When the ratio of RMF magnitude, B_ω, to external poloidal confinement field, B_e, is high, this resistivity is very adversely affected by the RMF drive process. However, when B_ω/B_e falls below about 0.3, the resistivity returns to values typical of non-driven FRCs. The observed scaling leads to a density dependence of n_e ∼ B_ω/r_sω^1/2 where r_s is the FRC separatrix radius and ω is the RMF frequency. Since the FRC contains little or no toroidal field, B_e is proportional to (n_eT_t)^1/2 where T_t = T_e + T_i is the sum of the electron and ion temperatures. In the present experiments, except for the initial start-up phase where T_t can exceed 100 eV, the plasma temperature is limited to about 40 eV by high oxygen impurity levels. Thus, low B_ω/B_e, low resistivity operation was only realized by operating at low values of B_ω. The RMF drive sustains particles as well as flux, and resistive input powers can be in the MW range at higher values of B_ω, so that high temperature, steady-state operation should be possible once impurity levels are reduced. Changes are being made to the present 'O-ring' sealed, quartz chambered TCS to provide bakable metal walls and wall conditioning as in other quasi-steady fusion facilities.

For the large helical device (LHD), the nonlinear evolution of equilibria that are linearly unstable to ideal interchange modes is studied using the reduced MHD equations. At sufficiently low beta, each individual mode saturates without affecting directly the evolution of the other modes. They only couple through the modification of the averaged pressure profile. The change of the averaged pressure profile is limited to the local flattening near the resonant surfaces. At higher beta values and for the same initial pressure profile, a bursting phenomenon in the kinetic energy is observed. This bursting activity is caused by the overlap of multiple modes, which results in a global reduction of the pressure. However, increasing beta and using a pressure profile obtained from the nonlinear evolution at the lower beta suppress this bursting behaviour. This result indicates that the pressure profile can be self-organized so that the LHD plasma could reach a high beta regime through a stable path.

Injection of cryogenic deuterium pellets has been successfully applied in ASDEX Upgrade for external edge localized mode (ELM) frequency control in type-I ELMy H-mode discharge scenarios. A pellet velocity of 560 m s⁻¹ and a size of about 6 × 10¹⁹ D-atoms was selected for technical reasons, although even lower masses were found sufficient to trigger ELMs. A moderate repetition rate close to 20 Hz was chosen to avoid over-fuelling of the core plasma. Pellet sequences of up to 4 s duration were injected into discharges close to the L–H threshold, intrinsically developing large compound ELMs at a rate of 3 Hz. With pellet injection, these large ELMs were completely replaced by smaller type-I ELMs at the much higher pellet frequency, accompanied by a slight increase of density and even of stored energy. This external ELM control could be repeatedly switched on and off by just interrupting the pellet train. ELMs were triggered in less than 200 µs after pellet arrival at the plasma edge, at which time only a fraction of the pellet has been ablated, forming a rather localized, three-dimensional plasmoid, which drives the edge unstable well before the deposited mass is spread toroidally. The pellet controlled case has also been compared with a discharge at a somewhat lower density, but with otherwise rather similar data, developing spontaneous 20 Hz type-I ELMs. Despite the different trigger mechanisms, the general ELM features turn out to be qualitatively similar, possibly because of the similarity of the two cases in terms of ELM relevant parameters. The scaling with background plasma, heating power, pellet launch parameters, etc over a larger range still remains to be investigated.

A new electromagnetic kinetic electron simulation model that uses a generalized split-weight scheme, where the adiabatic part is adjustable, along with a parallel canonical momentum formulation has been developed in three-dimensional toroidal flux-tube geometry. This model includes electron–ion collisional effects and has been linearly benchmarked. It is found that for H-mode parameters, the nonadiabatic effects of kinetic electrons increase linear growth rates of the ion-temperature-gradient-driven (ITG) modes, mainly due to trapped-electron drive. The ion heat transport is also increased from that obtained with adiabatic electrons. The linear behaviour of the zonal flow is not significantly affected by kinetic electrons. The ion heat transport decreases to below the adiabatic electron level when finite plasma β is included due to finite-β stabilization of the ITG modes. This work is being carried out using the 'Summit Framework'.

Neoclassical tearing modes (NTMs) are magnetohydrodynamic modes that can limit the performance of high beta discharges in tokamaks, in some cases leading to a major disruption. The destabilizing effect that results in NTM growth is a helical decrease in the bootstrap current caused by a local reduction of the plasma pressure gradient by 'seed' magnetic islands. The NTM is particularly well suited to control since the mode is linearly stable although nonlinearly unstable, so if the island amplitude can be decreased below a threshold size the mode will decay and vanish. One means of shrinking the island is the replacement of the 'missing' bootstrap current by a localized current generated by electron cyclotron current drive (ECCD). This method has been applied to the m = 3/n = 2 NTM in DIII-D, in H-mode plasmas with ongoing edge localized modes and sawteeth, both of which generate seed islands periodically. In the case of the 3/2 mode, full suppression was obtained robustly by applying about 1.5 MW of ECCD very near the rational surface of the mode. When the mode first appears in the plasma, the stored energy decreases by 30%, but after the mode is stabilized by the ECCD, the beta may be raised above the initial threshold pressure by 20% by additional neutral beam heating, thereby generating an improvement in the limiting beta of nearly a factor of 2. An innovative automated search algorithm was implemented to find and retain the optimum location for the ECCD in the presence of the mode.

A nonlinear instability due to zonal flows and magnetic islands has been found. The instability has the character of a dissipative drift instability due to an anomalous resistivity. The anomalous resistivity is typically two orders of magnitude larger than the classical at the edge.

Low frequency f = ⅕ − ⅓f_ci, (f_ci: ion gyro frequency in the external field B_w) waves are excited with an antenna which is compatible with a reactor in a plasma with field reversed configuration (FRC). Near and outside the separatrix r_s of the FRC plasma, though the applied wave is mainly in compressional mode, azimuthal and radial components are observed in the magnetic field disturbance of the excited wave, which propagate with the dispersion relation consistent with the shear Alfvén wave. These disturbances penetrate deep into the FRC plasma across the surface where the wave frequency exceeds local ion gyro frequency and propagate along magnetic lines of force with sound velocity, which behaviour is consistent with the shear Alfvén wave with finite temperature correction. Axial magnetic disturbance propagates axially and radially from the antenna across the plasma column.

There is currently both considerable interest in the physics of ELM transport in the scrape-off layer (SOL) and concern over the impact of the ELM power and particle loads on the divertor targets of future fusion reactors. This paper describes some experimental observations of the reaction of target floating potentials and currents during ELMs in TCV, relying principally on fast measurements of these parameters using tile embedded Langmuir probe arrays. Clear evidence is presented for rapid modifications in the local target floating potentials occurring long before the characteristic rise of hydrogenic excitation emission due to local recycling provoked by the arrival at the target of the ELM ion flux. This precursor activity appears to be synchronous with the growth of MHD modes in the main chamber. Simple conditional averaging is used to derive a 'coherent ELM' and thus generate the radial distribution of the current to probes held at the target potential. At some locations, the coherent ELM can also be used to estimate the time evolution of the local target electron temperature, density and power flux, even though these quantities are not directly measured. The time delays between the reactions of currents, floating potentials and derived temperature are consistent with the expections of recently published kinetic simulations of ELM energy transport down a one-dimensional SOL plasma. The strong potential variations observed during the ELM are the result of current flows at the targets. These currents are generally of opposite sign at inner and outer divertors and are thus consistent, at least in part, with a thermoelectric origin in which the driven current is produced by an in/out divertor temperature asymmetry near the strike point of nearly a factor 2, probably due to the generation of divertor asymmetries by the ELM heat pulse. Such asymmetries are commonly observed in low to medium density L-modes for the particular TCV magnetic equilibrium studied in this paper. The total current balance during the ELM is satisfied only to within a factor 2, so that, whilst some of the driven current flows parallel to field lines in the SOL, there is an apparent additional negative current to the inner divertor during the ELM whose origin remains unexplained. It might, however, be due, in part, to increased plasma–wall interaction in the main chamber during the ELM event.

The formation of internal transport barriers (ITBs) has been experimentally associated with the presence of rational q surfaces in both JET and ASDEX Upgrade. The triggering mechanisms are related to the occurrence of magneto-hydrodynamic (MHD) instabilities such as mode coupling and fishbone activity. These events could locally modify the poloidal velocity and increase transiently the shearing rate to values comparable with the linear growth rate of ion temperature gradient modes. For JET reversed magnetic shear scenarios, ITB emergence occurs preferentially when the minimum q reaches an integral value. In this case, transport effects localized in the vicinity of zero magnetic shear and close to rational q values may be at the origin of ITB formation. The role of rational q surfaces in ITB triggering stresses the importance of q profile control for an advanced tokamak scenario and could assist in substantially lowering the access power to these scenarios in next step facilities.

In the l = 3/m = 9 Uragan-3M (U-3M) torsatron (R₀ = 1 m, abar; ≈ 12 m, B_ϕ = 0.7 T, ι(abar)/2π ≈ 0.4), an open helical divertor has been realized. Recently, under RF plasma production and heating conditions, a strong up–down asymmetry of diverted plasma flow has been observed as a result of measurements of distributions of this flow in two symmetric poloidal cross-sections of the U-3M torus. In many aspects, this asymmetry is similar to that observed in the l = 2 Heliotron E (H-E) heliotron/torsatron under neutral beam injection and electron cyclotron heating conditions. The main feature of the asymmetry is a predominant outflow of the diverted plasma in the ion toroidal drift direction. On this basis, the asymmetry can be related to non-uniformity of the distribution of direct charged particle losses in the minor azimuth. In the work reported, the magnitude of diverted plasma flow in U-3M and the degree of its vertical asymmetry are studied as functions of the heating parameter , P being the RF power absorbed in the plasma, and are juxtaposed with corresponding P-related changes in the density, , and suprathermal ion content in the plasma. As the heating power increases, both the temperature of the main ion group and the relative content of suprathermal ions increase. At the same time, a decrease in plasma density is observed, evidencing a rise of particle loss. The rise of particle loss with heating could result from both a shift of diffusion regime towards a lower collisionality and a rise of direct (non-diffusive) loss of high-energy particles. Outside the confinement volume, the total flow of diverted plasma increases together with an increase of vertical flow asymmetry towards the ion toroidal drift side. Such a mutual accordance between the processes in the confinement volume and in the divertor region validates the hypothesis on a dominating role of fast particle loss in the formation of vertical asymmetry of divertor flow in U-3M. In conclusion, the results obtained on U-3M are compared with those from similar research on H-E.

The electron temperature gradient (ETG) driven instability in toroidal plasmas of parameter regimes close to stability boundaries is studied with gyrokinetic theory. Emphasis is placed on calculations of the maximum growth rate and threshold temperature gradient of the modes. Algebraic formulae for the maximum growth rate and for the critical gradient are presented. Estimations for electron thermal transport induced by ETG turbulence are formulated. The results are compared and shown to be in line with experimental observations.

A new divertor configuration (DIV-IIb) has been implemented in ASDEX Upgrade. In order to accommodate a large variety of plasma shapes with bottom triangularities (δ_bot) up to 0.48, the outer strikepoint region was modified and the roof baffle was lowered and diminished at its outer part in comparison with the previous divertor (DIV-II). The inner part of the divertor strikepoint module remains unchanged, but a smooth transition to the central column is provided at the divertor entrance to minimize local hydrogen recycling. An increase in power density is observed due to geometrical reasons at the outer target, whereas the divertor radiation for similar configurations and discharge conditions is unchanged. The pumping characteristics for D and He are almost retained, suggesting a large influence of the inner divertor leg, the configuration of which remains as before. Detachment in L-mode discharges fits well into a scaling deduced from JET data and earlier ASDEX Upgrade data. A significant reduction (20%) of the L–H threshold is observed compared with DIV-II. Its density dependence is weaker than in the previous DIV-II configuration and there are hints for an influence of triangularity on power threshold. Finally, clear evidence for a parasitic plasma below the divertor roof baffle is found.

The scrape-off layer (SOL) and divertor target plasma of a large spherical tokamak (ST) are characterized in detail for the first time. Scalings for the SOL heat flux width in MAST are developed and compared with conventional tokamaks. Modelling reveals the significance of the mirror force to the ST SOL. Core energy losses, including during ELMs, in MAST arrive predominantly (>80%) at the outboard targets in all geometries. Convective transport dominates energy losses during ELMs, and MHD analysis suggests ELMs in MAST are Type III even at auxiliary heating powers well above the L–H threshold. ELMs are associated with a poloidally and/or toroidally localized radial efflux at ∼1 km s⁻¹ well into the far SOL but not with any broadening of the target heat flux width. Toroidally asymmetric divertor biasing experiments have been conducted in an attempt to broaden the target heat flux width, with promising results. During vertical displacement events, the maximum product of the toroidal peaking factor and halo current fraction remains below 0.3, around half the ITER design limit. Evidence is presented to demonstrate that the resistance of the halo current path may have an impact on the distribution of the halo current.

Impurity injection in the JET ELMy H-mode regime has produced high-confinement, quasi-steady-state plasmas with densities close to the Greenwald density. However, at large Ar densities, a sudden loss of confinement is observed. A possible correlation between loss of confinement and the observed MHD phenomena, both in the core and in the edge of the plasma, was considered. The degradation in confinement coincided with impurity profile peaking following the disappearance of sawtooth activity. In addition, impurity density profile analysis confirmed that central MHD modes prevented impurity peaking. Experiments were designed to understand the role of sawtooth crashes in re-distributing impurities. Ion-cyclotron radio frequency heating was used to control the central q-profile and maintain sawtooth activity. This resulted in quasi-steady-state, high-performance plasmas with high Ar densities. At and high Ar injection rates, quasi-steady-states, which previously only lasted <1τ_E, were now maintained for the duration of the heating (Δ t ∼ 9τ _E). The increased central heating may have an additional beneficial effect in opposing impurity accumulation by changing the core power balance and modifying the impurity transport as predicted by neo-classical theory.

A new operational regime has been discovered on JFT-2M under the boronized first wall condition to produce the High Recycling Steady (HRS) H-mode, which is characterized by good energy confinement (H_89P ∼ 1.6) at a high density of around 70% of the Greenwald density (n_e/n_GW), low radiated power fraction, and complete disappearance of large ELMs. Accompanying the HRS H-mode transition, coherent magnetic fluctuation is seen on magnetic probes at the vessel wall. These coherent fluctuations have a frequency of the order of 10–100 kHz with significant variation. They are recognized as important in enhancing particle transport, whose characteristics are similar to the Enhanced D_α (EDA) H-mode regime reported from Alcator C-Mod. The H'-mode, previously observed on JFT-2M, had features in common with the EDA mode in terms of the coherent density fluctuation, but it appeared transiently. The HRS operating regime is also similar to the EDA mode, except that HRS is seen even at low q₉₅ (2 < q₉₅ ⩽ 3). The most important feature of the HRS H-mode edge condition is compatibility with an improved core confinement mode at a high density without large ELMs. We have demonstrated that an internal transport barrier can be produced under a HRS H-mode edge condition, achieving β_NH_89P ∼ 6.2 at n_e/n_GW ∼ 70%, transiently.

Theoretical studies aimed at predicting and diagnosing field-line quality in a spheromak are described. These include nonlinear three-dimensional MHD simulations and analyses of confinement in spheromaks dominated by either open (stochastic) field lines or approximate flux surfaces. Three-dimensional nonlinear MHD simulations confirm that field lines are predominantly open when there is a large-amplitude toroidal-mode-number n = 1 mode. However, an appreciable volume of good flux surfaces can be obtained either during the drive-off phase of a scheme with periodic pulsed drive or for an extended period under driven conditions, with oscillating volume, when the odd-n modes are suppressed. If a configuration with radially localized perturbations can be achieved, a scaling analysis for a Rosenbluth–Bussac spheromak equilibrium indicates a favourable (1/Lundquist number) scaling to larger, higher-field devices. A hyper-resistivity analysis, which also assumes small-scale perturbations, reproduces well magnetic probe data in the sustained spheromak physics experiment, while an analysis of the same experiment based on one-dimensional transport along open field lines contradicts experimental observations in several key ways. The scaling analysis is also applied to reversed-field pinches and indicates that a completely determined scaling can be obtained with less approximation to the resistive MHD equations than indicated in the previous literature.

The relationship between particle and heat transport in an internal transport barrier (ITB) has been systematically investigated in reversed shear (RS) and high β_p 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 (D_He) and the ion thermal diffusivity (χ_i) are 5–15 times higher than the neoclassical level in the high β_p mode plasma. In the RS plasma, D_He is reduced from 6–7 times to a 1.4–2 times higher level than the neoclassical level when χ_i 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 χ_i 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 β_p 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.

The Japanese reduced activation ferritic steels (RAFSs) R&D road map towards DEMO is shown. The important steps include high-dose irradiation in fission reactors such as the high flux isotope reactor at Oak Ridge National Laboratory, irradiation tests with 14 MeV neutrons in the International Fusion Materials Irradiation Facility and application to ITER test blanket modules to provide an adequate database of RAFSs for the design of DEMO. The current status of RAFS development is also introduced. The major properties of concern are well-known, and process technologies are mostly ready for fusion application. RAFSs are now certainly ready to proceed to the next stage. A materials database is already in hand, and further progress is anticipated with the design of the ITER test blanket. Oxide dispersion strengthening steels are quite promising for high temperature operation of the blanket system, with potential improvements in radiation resistance and in corrosion resistance.

The edge plasma density at the high-field side (HFS) and low-field side (LFS) was measured using a FIR interferometer and an O-mode reflectometer simultaneously in order to study the asymmetry of collapse of the density pedestal by type I ELM. From fast time series analysis using the same digitizer system for the reflectometer, D_α intensity, interferometer and magnetic probe, experimental evidence for poloidal asymmetry of the collapse was clearly observed in JT-60U ELMy H-mode discharges for the first time. The collapse of the pedestal density is manifested as a fast inward displacement of the cutoff layer at the LFS, the increase of D_α intensity and large magnetic oscillations. No immediate response, however, was observed in the line-integrated density at the HFS of the plasma (FIR-U1). Instead, the signal showed a small reduction after a time delay of about L_∥/C_s, which was attributed to the parallel flow from the inside to outside in order to fill the density gap as a result of localized particle expulsion from the LFS. Furthermore, no density increase in the HFS scrape-off layer (SOL) was observed during the horizontal plasma sweep experiments corresponding to particles expelled from the HFS midplane to the SOL directly, at the onset of density collapse and before the increase of D_α intensity. Together, these results indicate that the collapse of the density pedestal during type I ELMs is localized at the LFS midplane.

Bursting modes, which are identified as bounce precession frequency fishbone modes, are observed on a low aspect ratio tokamak, the national spherical torus experiment (Ono M. et al 2000 Nucl. Fusion40 557). They are predicted to be important in high current, low shear discharges with a significant population of trapped particles with a large mean bounce angle, such as produced by near tangential beam injection into a small aspect ratio device. Such a distribution is often stable to the usual precession-resonance fishbone mode. These modes could be important in ignited plasmas, driven by the trapped alpha particle population.

The behaviour of the density profiles in ASDEX Upgrade can be described well with the assumption D ∝ χ and a pinch of the order of the neoclassical Ware pinch. The latter is estimated from slowly equilibrating density profiles. The assumption D ∝ χ has been succesfully tested by varying the heat deposition profile, making use of on-/off-axis ICRH and ECRH: due to the generally observed self-similarity of the temperature profile, such variations in the heat flux profile have a strong effect on the χ-profile and on the D-profile if the above assumption is correct. The corresponding variations in the density profiles have indeed been observed. The model is also capable of describing the decay of the density profile after injecting a train of pellets. The anomalous transport of impurities is also increased with central heating, and the corresponding flattening of the density profile leads to a significant reduction of the neoclassical impurity pinch. Central ICR or ECR heating are therefore now routinely used to control density peaking and its negative effect on the stability of neoclassical tearing modes as well as to control the impurity transport in ASDEX Upgrade.

This paper reports results on the progress in steady-state high-β_p ELMy H-mode discharges in JT-60U. A fusion triple product, n_D(0)τ_ET_i(0), of 3.1 × 10²⁰ m⁻³ s keV under full non-inductive current drive has been achieved at I_p = 1.8 MA, which extends the record value of the fusion triple product under full non-inductive current drive by 50%. A high-beta plasma with β_N ∼ 2.7 has been sustained for 7.4 s (∼60τ_E), with the duration determined only by the facility limits, such as the capability of the poloidal field coils and the upper limit on the duration of injection of neutral beams. Destabilization of neoclassical tearing modes (NTMs) has been avoided with good reproducibility by tailoring the current and pressure profiles. On the other hand, a real-time NTM stabilization system has been developed where detection of the centre of the magnetic island and optimization of the injection angle of the electron cyclotron wave are done in real time. By applying this system, a 3/2 NTM has been completely stabilized in a high-beta region (β_p ∼ 1.2, β_N ∼ 1.5), and the beta value and confinement enhancement factor have been improved by the stabilization.

Significant progress in obtaining high-performance discharges under quasi-steady-states in the HT-7 superconducting tokamak has been realized since the last IAEA meeting. In relation to the previous experiments, various features of the non-inductive current driven, heating, profile control, MHD stabilization and edge physics are integrated and optimized to achieve steady-state high-performance discharges. Both on-axis and off-axis electron heating with global peaked and locally steepened electron pressure profiles were realized with improved confinement if the ion Bernstein wave (IBW) resonant layer was properly selected. Stabilization of MHD instabilities was demonstrated by off-axis IBW heating. The internal transport barrier structure was formed by off-axis lower hybrid current drive (LHCD). Long-pulse discharges with T_e ∼ 1 keV and central density ∼1 × 10¹⁹ m⁻³ were obtained with a duration of 10–20 s. A combination of IBW heating and LHCD produced a broadened current density profile, which may be a signature of the synergy effect between two waves. Experimental results show that features of IBW in controlling electron pressure profile can be integrated into LHCD target plasmas. HT-7 has produced a variety of discharges with β_N × H₈₉ > 1–4 for durations of several to several tens of energy confinement times with a non-inductive driven current of 50–80% by optimizing the IBW heating and LHCD and avoiding MHD activities.

Compatibility between the plasma and low activation ferritic steel, which is a candidate material for fusion demonstration reactors, has been investigated step by step in the JFT-2M tokamak. We have entered the third stage of the Advanced Material Tokamak EXperiment (AMTEX), where the inside of the vacuum vessel wall is completely covered with ferritic steel plates ferritic inside wall (FIW). The effects of a FIW on the plasma production, impurity release, the operation region, and H-mode characteristics have been investigated. No negative effect has been observed up to now. A high normalized beta plasma of β_N ∼ 3, having both an internal transport barrier and a steady H-mode edge was obtained. A remarkable reduction in ripple trapped loss from 0.26 MW m⁻² (without ferritic steel) to less than 0.01 MW m⁻² was demonstrated by the optimization of the thickness profile of FIW. A code to calculate fast ion losses, taking into account the full three-dimensional magnetic structure was developed, and values obtained using the code showed good agreement with experimental results. Thus, encouraging results are obtained for the use of this material in the demo-reactor.