Significant progress has been made in the area of advanced modes of operation that are candidates for achieving steady state conditions in a fusion reactor. The corresponding parameters, domain of operation, scenarios and integration issues of advanced scenarios are discussed in this chapter. A review of the presently developed scenarios, including discussions on operational space, is given. Significant progress has been made in the domain of heating and current drive in recent years, especially in the domain of off-axis current drive, which is essential for the achievement of the required current profile. The actuators for heating and current drive that are necessary to produce and control the advanced tokamak discharges are discussed, including modelling and predictions for ITER. The specific control issues for steady state operation are discussed, including the already existing experimental results as well as the various strategies and needs (qψ profile control and temperature gradients). Achievable parameters for the ITER steady state and hybrid scenarios with foreseen heating and current drive systems are discussed using modelling including actuators, allowing an assessment of achievable current profiles. Finally, a summary is given in the last section including outstanding issues and recommendations for further research and development.

A practical method for performing a tokamak equilibrium reconstruction in real time for arbitrary time-varying discharge shapes and current profiles is described. An approximate solution to the Grad-Shafranov equilibrium relation is found which best fits the diagnostic measurements. Thus a solution for the spatial distribution of poloidal flux and toroidal current density is available in real time that is consistent with plasma force balance, allowing accurate evaluation of parameters such as discharge shape and safety factor profile. The equilibrium solutions are produced at a rate sufficient for discharge control. This equilibrium reconstruction algorithm has been implemented on the digital plasma control system for the DIII-D tokamak. The first application of a real time equilibrium reconstruction to discharge shape control is described.

No abstract

No abstract

Poloidal and toroidal rotation of the main ions (He2+) and the impurity ions (C6+ and Bs+) in Hmode helium plasmas have been measured via charge exchange recombination spectroscopy in the DIII-D tokamak. It was discovered that the main ion poloidal rotation is in the ion diamagnetic drift direction while the impurity ion rotation is in the electron diamagnetic drift direction, in qualitative agreement with the neoclassical theory. The deduced radial electric field in the edge is of the same negativewell shape regardless of which ion species is used, validating the fundamental nature of the electric field in L-H transition phenomenology.PACS numbers: 52.55.Fa, 52.55.Pi, 52.70.Kz Since its original discovery in ASDEX [1],the H mode has proven to be one of the most robust and ubiquitous modes of improved confinement in toroidal magnetic fusion devices. The physics of the L-mode to H-mode (L-H) transition has attracted a great deal of interest and effort from both the experimental and theoretical communities. Critical reviews of this subject have recently been made by Burrell et al. [2] and Groebner [3]. Although a complete, quantitative theory of the confinement improvement at the L-H transition does not yet exist, combined theoretical and experimental work has merged into a paradigm that a highly sheared EXB flow in the plasma edge can lead to better confinement through decorrelation of the fluctuations, decreased radial correlation lengths, and reduced turbulent transport [4,5]. The experimental observations that the edge impurity ion poloidal rotation [6-9] and the edge radial electric field [6-11] change dramatically and abruptly at the L-H transition have led to several theories which consider how the radial electric field or the (main) ion poloidal rotation changes across the L-H transition. Among these are theories based on bifurcation of the radial electric field [5,12], Stringer spin-up [13,14], turbulent Reynolds stress [15], temperature gradient-induced poloidal rotation [16], and particle and energy confinement bifurcation [17].Only the impurity ion rotation measurements have been available previously and some theoretical models were developed to explain the sudden (~I msec) increase of poloidal rotation (in the electron diamagnetic drift direction) at the L-H transition; these theories implicitly assumed that the main ion rotation and the impurity ion rotation are identical. However, recent neoclassical derivations of rotation velocities predict that the main ion poloidal rotation and the impurity ion poloidal rotation speed could be quite different [18,19]. Although not in H mode, a comparison of the main and the impurity ion poloidal rotation in Ohmic plasmas in TEXT was reported previously [20], which showed little diff'erence between them. Experimental clari6cation of the rotation behavior of the main and impurity iona in H-mode plasmas is important not only as a check of the neoclassical rotation theories, but also as a test of the existing L-H transition theories. However, charge exchange recom...