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A recently proposed paradigm for understanding turbulent transport dynamics in magnetically confined plasmas, based on the concept of self-organized criticality (SOC), is extended to include the interplay among the many transport mechanisms existing in real plasmas. This extended model might shed some light on the experimentally observed violation at fluctuation scales of the scale invariance condition inherent to the standard SOC model. At the same time, it might provide new experimental tests that could help to validate the possible relevance of the SOC model for understanding turbulent plasma dynamics, since some computations based on a cellular automata realization of such an extended model reveal a dramatic change in the dynamics, with the edge coming to play a dominant role. This might give some hints towards understanding the experimental fact relating good plasma core confinement to a good conditioning of the edge within an SOC paradigm.

In the JFT-2M tokamak, the application of low activation ferritic steels to plasmas has been investigated (the so-called Advanced Material Tokamak Experiment (AMTEX) programme). In the first stage, toroidal field ripple reduction was examined by ferritic steel boards (FBs) inserted between the toroidal field coils and the vacuum vessel. It is demonstrated that FB insertion reduced toroidal field ripple and the losses of fast ions produced by tangential co-NBI. By optimizing the FB thickness, such that fundamental mode ripple is minimized to 0.07% at the shoulder of the inside wall, ripple trapped loss is reduced to an almost negligible level. It is determined that the reductions of the fundamental mode ripple and the ripple banana diffusion coefficients at the shoulder are most effective in reducing ripple ion losses. Ripple loss reduction by FBs is also confirmed with perpendicular beam injection. The insertion of FBs causes no undesirable effects on plasma production and control.

A configuration in which the vacuum vessel (VV) fits tightly to the plasma aids the passive plasma vertical stability, and ferromagnetic material in the VV reduces the toroidal field ripple. The blanket modules are supported directly by the VV. A full scale VV sector model has provided critical information related to fabrication technology and for testing the magnitude of welding distortions and achievable tolerances. This R&D validated the fundamental feasibility of the double wall VV design. The blanket module configuration consists of a shield body to which a separate first wall is mounted. The separate first wall has a facet geometry consisting of multiple flat panels, where 3-D machining will not be required. A configuration with deep slits minimizes the induced eddy currents and loads. The feasibility and robustness of solid hot isostatic pressing joining were demonstrated in the R&D by manufacturing and testing several small and medium scale mock-ups and finally two prototypes. Remote handling tests and assembly tests of a blanket module have demonstrated the basic feasibility of its installation and removal.

Co-current central impurity toroidal rotation has been observed in Alcator C-Mod plasmas with on-axis ICRF heating. The rotation velocity increases with plasma stored energy and decreases with plasma current. Very similar behaviour has been seen during ohmic H modes, which suggests that the rotation, generated in the absence of an external momentum source, is not mainly an ICRF effect. A scan of ICRF resonance location across the plasma has been performed in order to investigate possible influences on the toroidal rotation. With a slight reduction of toroidal magnetic field from 4.7 to 4.5 T and a corresponding shift of the ICRF resonance from r/a = -0.36 to -0.48, the central toroidal rotation significantly decreased together with the formation of an internal transport barrier (ITB). During the ITB phase, electrons and impurities peaked continuously for |r/a| ⩽ 0.5. Comparison of the observed rotation and neoclassical predictions indicates that the core radial electric field changes from positive to negative during the ITB phase. Similar rotation suppression and ITB formation have been observed during some ohmic H mode discharges.

With use of a multimachine pedestal database, essential issues for each regime of ELM types are investigated. They include (i) understanding and prediction of pedestal pressure during type I ELMs, a reference operation mode of a future tokamak reactor; (ii) identification of the operation regime of type II ELMs, which have small ELM amplitude with good confinement characteristics; (iii) identification of the upper stability boundary of type III ELMs for access to the higher confinement regimes with type I or II ELMs; (iv) understanding the relation between core confinement and pedestal temperature in conjunction with the confinement degradation in high density discharges. Both scaling and model based approaches for expressing pedestal pressure are shown to roughly scale the experimental data well and could be used to make initial predictions for a future reactor. q and δ are identified as important parameters for obtaining the type II ELM regime. A theoretical model of type III ELMs is shown to reproduce the upper stability boundary reasonably well. It is shown that there exists some critical pedestal temperature below which the core confinement starts to degrade. It is also shown that it is possible to obtain improved pedestal conditions for good confinement in high density discharges by increasing the plasma triangularity.

The dependence of edge stability on plasma shape and local pressure gradients P' in the DIII-D and JT-60U tokamaks is studied. The stronger plasma shaping in DIII-D allows the edge region of DIII-D discharges with type I (giant) ELMs to have access to the second region of stability for ideal ballooning modes and a larger edge pedestal pressure gradient P' than JT-60U type I ELM discharges. These JT-60U discharges are near the ballooning mode first regime stability limit. The DIII-D results support an ideal stability based working model of type I ELMs as low to intermediate toroidal mode number, n, MHD modes. The results from a stability analysis of JT-60U type I ELM discharges indicate that the predictions of this model are also consistent with JT-60U edge stability observations.

A plasma confined in a dipole field exhibits unique equilibrium and stability properties. In particular, equilibria exist at all beta values and these equilibria are found to be stable to ballooning modes when they are interchange stable. When a kinetic treatment is performed at low beta, a drift temperature gradient mode is also found which couples to the MHD mode in the vicinity of marginal interchange stability.

ITER is planned to be the first fusion experimental reactor in the world operating for research in physics and engineering. The first ten years of operation will be devoted primarily to physics issues at low neutron fluence and the following ten years of operation to engineering testing at higher fluence. ITER can accommodate various plasma configurations and plasma operation modes, such as inductive high Q modes, long pulse hybrid modes and non-inductive steady state modes, with large ranges of plasma current, density, beta and fusion power, and with various heating and current drive methods. This flexibility will provide an advantage for coping with uncertainties in the physics database, in studying burning plasmas, in introducing advanced features and in optimizing the plasma performance for the different programme objectives. Remote sites will be able to participate in the ITER experiment. This concept will provide an advantage not only in operating ITER for 24 hours a day but also in involving the worldwide fusion community and in promoting scientific competition among the ITER Parties.

External impurity injection into L mode edge discharges in DIII-D has produced clear confinement improvement (a factor of 2 in energy confinement and neutron emission), reduction in all transport channels (particularly ion thermal diffusivity to the neoclassical level), and simultaneous reduction of long wavelength turbulence. Suppression of the long wavelength turbulence and transport reduction are attributed to synergistic effects of impurity induced enhancement of E × B shearing rate and reduction of toroidal drift wave turbulence growth rate. A prompt reduction of density fluctuations and local transport at the beginning of impurity injection appears to result from an increased gradient of toroidal rotation enhancing the E × B shearing. Transport simulations carried out using the National Transport Code Collaboration demonstration code with a gyro-Landau fluid model, GLF23, indicate that E × B shearing suppression is the dominant transport suppression mechanism.

An ion cyclotron range of frequency (ICRF) heating experiment was conducted in the third campaign of LHD in 1999. 1.35 MW of ICRF power were injected into the plasma and 200 kJ of stored energy were obtained, which was maintained for 5 s by ICRF power only after the termination of ECH. The impurity problem was so completely overcome that the pulse length was easily extended to 68 s at a power level of 0.7 MW. The utility of a liquid stub tuner in steady state plasma heating was demonstrated in this discharge. The energy confinement time of the ICRF heated plasma has the same dependences on plasma parameters as those of the ISS95 stellarator scaling with a multiplication factor of 1.5, which is a high efficiency comparable to that of NBI. Such an improvement in performance was obtained by various means, including: (a) scanning of the magnetic field intensity and minority concentration, (b) improvement of particle orbits due to a shift of magnetic axis and (c) reduction of the number of impurity ions by means of titanium gettering and the use of carbon divertor plates. In the optimized heating regime, ion heating turned out to be the dominant heating mechanism, unlike in CHS and W7-AS. Owing to the high quality of the heating and the parameter range being extended far beyond that of previous experiments, the experiment can be regarded as the first complete demonstration of ICRF heating in stellarators.

A series of fusion neutronics experiments has been performed at the Fusion Neutronics Source facility at Japan Atomic Energy Research Institute as ITER EDA R&D tasks to find ways of dealing with various nuclear problems originating from 14 MeV neutrons in ITER. Recently three groups of experiments were carried out: (1) straight duct streaming experiments, (2) decay heat experiments, and (3) development of a fusion power monitor utilizing activation of water. The straight duct streaming experiments suggest that the calculation accuracy for straight duct streaming analyses in ITER nuclear designs is ±40%. The decay heat experiments show that the accuracy of the decay heat calculation is within 10% for copper and type 316 stainless steel, while it is ∼30% for tungsten. It is demonstrated that a fusion power monitor utilizing activation of water is applicable to ITER.

Experimental devices for the study of the physics of high beta (β ≳ 4%), low aspect ratio (A ≲ 4.5) stellarator plasmas require coils that will produce plasmas satisfying a set of physics goals, provide experimental flexibility and be practical to construct. In the course of designing a flexible coil set for the National Compact Stellarator Experiment, several innovations have been made that may be useful in future stellarator design efforts. These include: the use of singular value decomposition methods for obtaining families of smooth current potentials on distant coil winding surfaces from which low current density solutions may be identified; the use of a control matrix method for identifying which few of the many detailed elements of a stellarator boundary must be targeted if a coil set is to provide fields to control the essential physics of the plasma; the use of a genetic algorithm for choosing an optimal set of discrete coils from a continuum of potential contours; the evaluation of alternate coil topologies for balancing the trade-off between physics objectives and engineering constraints; the development of a new coil optimization code for designing modular coils and the identification of a `natural' basis for describing current sheet distributions.