[en] After a general discussion of the experimental characteristics of the L-H transition and consideration of basic theoretical principles underlying models for it, this paper reviews the various theories of the L-H transition available in the literature, providing some background information on each theory and expressing the transition criteria in forms suitable for comparison with experiment. Some conclusions on the relevance of these models for explaining the experimental data on the transition are drawn. (author)

[en] The dependences of heat transport on the dimensionless plasma physics parameters has been measured for both L-mode and H-mode plasmas on the DIII-D tokamak. Heat transport in L-mode plasmas has a gyroradius scaling that is gyro-Bohm-like for electrons and worse than Bohm-like for ions, with no measurable beta or collisionality dependence; this corresponds to having an energy confinement time that scales like τ_E ∝ n^0.5 P^-0.5. H-mode plasmas have gyro-Bohm-like scaling of heat transport for both electrons and ions, weak beta scaling, and moderate collisionality scaling. In addition, H-mode plasmas have a strong safety factor scaling (χ ∼ q²) at all radii. Combining these four dimensionless parameter scalings together gives an energy confinement time scaling for H-mode plasmas like τ_E ∝ B^-1 ρ^-3.15 β^0.03 ν^-0.42 q₉₅^-1.43 ∝ I^0.84 B^0.39 n^0.18 P^-0.41 L^2.0, which is similar to empirical scalings derived from global confinement databases. (author)

[en] Transport barriers and L-H transitions in Tokamak plasmas are often attributed to suppression of turbulence by a shear flow related to a plasma gradient, eg. of density. However, such shear flow is also affected by the second derivative of density. When this is introduced there is no unique relation between flux and gradient - it depends on the source distribution within the plasma and on conditions at the plasma edge (eg. imposed by the scrape-off layer). This edge gradient must lie within prescribed limits if a stationary plasma profile (which may include an improved confinement zone) is to exist. (author)

[en] The plasma performance in two design options of the reduced-technical objectives/reduced cost (RTO/RC) ITER, i.e. IAM (intermediate aspect ratio machine) and LAM (low aspect ratio machine) is analysed. It is shown that Q=P_fus/P_aux∼10 can be obtained in both options at inductively driven ELMy H-mode operation. The operation domain in LAM is found to be marginally larger than that in IAM. The non-inductive operation with Q approx.= 5 will be possible in both machines, provided a large amount of power with a high current drive efficiency is applied, or substantial improvement of the energy confinement time relative to the ELMy H-mode (H_H=1.2-1.4) is obtained. The required values of H_H and β_N are marginally smaller in IAM. The IAM-like machine, ITER-FEAT (fusion energy advanced tokamak), proposed for a detailed engineering design is discussed in brief. (author)

[en] The thickness of the transport barrier at the plasma edge is discussed, by analyzing the structure of the interface region between the turbulent L-mode region and the region where the transport is strongly-stabilized by the electric field. The effect of this localized radial electric field is prescribed. The spatial profile of turbulence intensity is analyzed by using a simplified model, in which the suppression and transport of turbulence intensity are introduced. The scaling property of the transport barrier is discussed. (paper)

[en] The ITER Physics Basis presents and evaluates the physics rules and methodologies for plasma performance projections, which provide the basis for the design of a tokamak burning plasma device whose goal is to demonstrate the scientific and technological feasibility of fusion energy for peaceful purposes. This Chapter summarizes the physics basis for burning plasma projections, which is developed in detail by the ITER Physics Expert Groups in subsequent chapters. To set context, the design guidelines and requirements established in the report of ITER Special Working Group 1 are presented, as are the specifics of the tokamak design developed in the Final Design Report of the ITER Engineering Design Activities, which exemplifies burning tokamak plasma experiments. The behaviour of a tokamak plasma is determined by the interaction of many diverse physics processes, all of which bear on projections for both a burning plasma experiment and an eventual tokamak reactor. Key processes summarized here are energy and particle confinement and the H-mode power threshold; MHD stability, including pressure and density limits, neoclassical islands, error fields, disruptions, sawteeth, and ELMs; power and particle exhaust, involving divertor power dispersal, helium exhaust, fuelling and density control, H-mode edge transition region, erosion of plasma facing components, tritium retention; energetic particle physics; auxiliary power physics; and the physics of plasma diagnostics. Summaries of projection methodologies, together with estimates of their attendant uncertainties, are presented in each of these areas. Since each physics element has its own scaling properties, an integrated experimental demonstration of the balance between the combined processes which obtains in a reactor plasma is inaccessible to contemporary experimental facilities: it requires a reactor scale device. It is argued, moreover, that a burning plasma experiment can be sufficiently flexible to permit operation in a steady state mode, with non-inductive plasma current drive, as well as in a pulsed mode where current is inductively driven. Overall, the ITER Physics Basis can support a range of candidate designs for a tokamak burning plasma facility. For each design, there will remain a significant uncertainty in the projected performance, but the projection methodologies outlined here do suffice to specify the major parameters of such a facility and form the basis for assuring that its phased operation will return sufficient information to design a prototype commercial fusion power reactor, thus fulfilling the goal of the ITER project. (author)

No abstract available

[en] In magnetically confined plasmas there is evidence of localized regions of improved confinement. These regions are usually associated with shear flows with radial structure, and an important problem is to understand how such flows emerge. To address this problem a reaction-diffusion type model of turbulence-shear flow interaction that incorporates the mechanism of turbulence suppression by shear, and parameterizes turbulent transport as a nonlinear diffusivity is considered. The fixed points of the model correspond to the L (low confinement) and H (high confinement) modes of the system, and it is shown that for a range of parameter values the H-mode fixed point has a finite-k instability. Numerical results show that this instability leads, in the nonlinear regime, to the formation of stratified shear layers and jets in which bands of intense shear and suppressed turbulence alternate with bands of low shear and enhanced turbulence. Approximate analytical solutions of the model corresponding to high-confinement modes with radial structure are presented

[en] H-mode operational boundaries and H-mode confinement are investigated on ASDEX Upgrade. The local edge parameter threshold for H-mode holds independent of divertor geometry and changes little with ion mass. The deviation of the H-mode power threshold at densities near the Greenwald limit can be understood as a consequence of a confinement deterioration, caused by 'stiff' temperature profiles and lack of core density gradients in gas puff fuelled discharges. Ion and electron temperature profiles can be described by a lower limit of gradient length L_T=T/T'. (author)

[en] ITER will require a level of energy (τ_E) and particle (τ_p) confinement sufficient for reaching ignition and extended burn of DT plasmas, with steady-state, high-Q (Q = P_fusion/P_aux = 5P_α/P_aux) operation as an ultimate goal. This translates into a required confinement capability of T_i(0)n_DT(0)τ_E ∼4--8 x 10²¹ keV·s/m³ for Q ∼5--∞ [where T_i(0) is the central ion temperature, n_DT(0) is the central DT fuel ion density, τ_E is the global energy confinement time]. In addition, τ_p^He/τ_E < 10 is required to ensure that the thermal α-particle (He) accumulation (n_He/n_e) is less than 10% (where τ_p^He is the global thermal alpha particle containment time). Furthermore, the tritium fuel burnup fraction under nominal burn conditions must be greater than 1.5%. While the ITER Tnτ_E requirement is about ten times the highest value achieved in JET [3], the extrapolation in plasma physics parameters is less. If β (ratio of plasma kinetic pressure to magnetic pressure), ν* (collisionality), and ρ/a (ratio of the ion gyroradius to the plasma minor radius) are taken as the relevant dimensionless parameters to characterize the plasma, the extrapolation from present JET performance to ITER is a factor ≤2. The ITER concept is based on the expectation that H-mode confinement can be achieved for long pulse. This paper covers the energy and particle confinement issues for ITER. 15 refs