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The understanding and predictive capability of transport physics and plasma confinement is reviewed from the perspective of achieving reactor-scale burning plasmas in the ITER tokamak, for both core and edge plasma regions. Very considerable progress has been made in understanding, controlling and predicting tokamak transport across a wide variety of plasma conditions and regimes since the publication of the ITER Physics Basis (IPB) document (1999 Nucl. Fusion 39 2137-2664). Major areas of progress considered here follow. (1) Substantial improvement in the physics content, capability and reliability of transport simulation and modelling codes, leading to much increased theory/experiment interaction as these codes are increasingly used to interpret and predict experiment. (2) Remarkable progress has been made in developing and understanding regimes of improved core confinement. Internal transport barriers and other forms of reduced core transport are now routinely obtained in all the leading tokamak devices worldwide. (3) The importance of controlling the H-mode edge pedestal is now generally recognized. Substantial progress has been made in extending high confinement H-mode operation to the Greenwald density, the demonstration of Type I ELM mitigation and control techniques and systematic explanation of Type I ELM stability. Theory-based predictive capability has also shown progress by integrating the plasma and neutral transport with MHD stability. (4) Transport projections to ITER are now made using three complementary approaches: empirical or global scaling, theory-based transport modelling and dimensionless parameter scaling (previously, empirical scaling was the dominant approach). For the ITER base case or the reference scenario of conventional ELMy H-mode operation, all three techniques predict that ITER will have sufficient confinement to meet its design target of Q = 10 operation, within similar uncertainties.

Effects of plasma rotation and losses of fast ions on Type-I ELM characteristics have been systematically studied in the JT-60U tokamak, scanning combinations of NBI (tangential co-, balanced- and counter-NBI plus perpendicular NBI) in the three types of plasma configurations (corresponding toroidal field ripple at the plasma edge, δ r ∼0.4, 1.0 and 2.0%). New findings on the Type-I ELM characteristics are as follows: smaller ELM energy loss normalized by pedestal stored energy, ΔW ELM /W ped , and faster ELM frequency, f ELM , are confirmed in the counter-NBI in comparison with the co-NBI discharges. Nevertheless, the power loss due to ELM, P ELM (= ΔW ELM × f ELM ), normalized by heating power crossing the separatrix, P SEP , is constant regardless of the direction of the momentum injection at each plasma configuration. In contrast, the P ELM /P SEP decreases with increasing δ r . The relationship between ELMs, rotation and the losses of fast ions are discussed.

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.

Advanced tokamak regimes obtained in ASDEX Upgrade, DIII-D, FT-U, JET, JT-60U, TCV and Tore Supra experiments are assessed both in terms of their fusion performance and capability for ultimately reaching steady-state using data from the international internal transport barrier database. These advanced modes of tokamak operation are characterized by an improved core confinement and a modified current profile compared to the relaxed Ohmically driven one. The present results obtained in these experiments are studied in view of their prospect for achieving either long pulses ('hybrid' scenario with inductive and non-inductive current drive) or ultimately steady-state purely non-inductive current drive operation in next step devices such as ITER. A new operational diagram for advanced tokamak operation is proposed where the figure of merit characterizing the fusion performances and confinement, H × βN / q 295, is drawn versus the fraction of the plasma current driven by the bootstrap effect. In this diagram, present day advanced tokamak regimes have now reached an operational domain that is required in the non-inductive ITER current drive operation with typically 50% of the plasma current driven by the bootstrap effect (Green et al 2003 Plasma Phys. Control. Fusion 45 587). In addition, the existence domain of the advanced mode regimes is also mapped in terms of dimensionless plasmas physics quantities such as normalized Larmor radius, normalized collisionality, Mach number and ratio of ion to electron temperature. The gap between present day and future advanced tokamak experiments is quantitatively assessed in terms of these dimensionless parameters.

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.

The ITER Threshold Database, which at present comprises data from nine divertor tokamaks, is described. The main results are presented and discussed. The properties and dependences of the power threshold in individual devices are reviewed. In particular, the analysis shows a rather general linear dependence on magnetic field, but a non-monotonic density dependence that varies from device to device. Investigation of the combined database suggests that the threshold dependence Pthres approximately=0.3neBT2.5 shows reasonable agreement with the data. This expression yields Pthres approximately=150 MW at a density of 0.5*1020 m-3 for ITER. Other expressions with weaker size dependence, and therefore lower threshold power for ITER, are also discussed. Their agreement with the present data is poorer than that of the above expression. In addition, the database is investigated by statistical discriminant analysis. The edge data included at present are described and discussed. Finally, there is a discussion of the implications of the results for ITER

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.

The Fusion Experimental Reactor (FER) is a D-T-burning tokamak machine currently being designed. It is expected to provide physical information and technical experiences that will be sufficient to proceed towards the DEMO Fusion Reactor which will demonstrate electric power generation by fusion energy. An efficient ash exhaust, a hybrid current drive operation, the use of a 3% ripple field, the technological achievements in R&D of the magnets, and the negative-ion beam system are expected to allow the FER to achieve its cost-effectiveness. >

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 study of a steady-state tokamak, JT-60 Super Upgrade, is being carried out. The capability of the present JT-60 facility will be fully utilized for this upgrade. The mission of JT-60SU is to establish integrated basis of physics and technology for steady-state tokamak reactors. In JT-60SU, steady-state physics will be evaluated in the intermediate parameter region between the present tokamaks and steady-state tokamak reactors. Technology development for long pulse operation and research for engineering safety will also be pursued.

On the basis of a high bootstrap current fraction observation with JT-60, the concept of a Steady-State Tokamak Reactor, the SSTR, was conceived and was developed with the design activity of the SSTR at JAERI (Japan Atomic Energy Research Institue). Results of ITER/FER (International Thermonuclear Experimental Reactor/Fusion Experimental Reactor) design activities have enhanced the SSTR design. Moreover, the progress of R&D for fusion reactor engineering, especially in the development of superconducting coils and negative-ion-based NBI at JAERI, has promoted the SSTR conceptual design as a realistic power reactor. Although present fusion power reactor designs are currently considered to be too large and costly, results of the SSTR conceptual design suggest that an efficient and promising tokamak reactor will be feasible. The conceptual design of the SSTR provides a realistic reference for a demo tokamak reactor. Notably, a reduction of the circulating power of the power station facility, owing to spontaneous bootstrap current, may ensure the high Q around 50 necessary for a commercial power reactor. >

The physics of ITER (International Thermonuclear Experimental Reactor) is based on demonstrated tokamak physics and credible extrapolations of that physics. Assessment of the energy confinement and MHD stability requirements led to the choice of the major plasma parameters of 22 MA for the plasma current, a toroidal field of approximately 5 T, an aspect ratio of approximately 3, and an elongation of approximately 2. Among the major accomplishments of the physics group has been the development of a database and an empirical scaling for L-mode energy confinement and the facilitation of an H-mode database and scaling. The divertor heat loads have been estimated by using experimentally validated models. The thermal and mechanical loads due to off-normal events such as disruptions have been based on analysis of the data from other tokamaks. To achieve the required availability of 10%, the pulse length has been extended by the use of current drive using 75 MW 1.3 MeV neutral beam and 45 MW lower hybrid systems. A relatively complete set of plasma diagnostics is planned for ITER. Finally, a physics R and D program has been developed. >

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.