Core micro-instability analysis of JET hybrid and baseline discharges with carbon wall

The core micro-instability characteristics of hybrid and baseline plasmas in a selected set of JET plasmas with carbon wall are investigated through local linear and non-linear and global linear gyro-kinetic simulations with the GYRO code [J. Candy and E. Belli, General Atomics Report GA-A26818 (2011)]. In particular, we study the role of plasma pressure on the micro-instabilities, and scan the parameter space for the important plasma parameters responsible for the onset and stabilization of the modes under experimental conditions. We find that a good core confinement due to strong stabilization of the micro-turbulence driven transport can be expected in the hybrid plasmas due to the stabilizing effect of the fast ion pressure that is more effective at the low magnetic shear of the hybrid discharges. While parallel velocity gradient destabilization is important for the inner core, at outer radii the hybrid plasmas may benefit from a strong quench of the turbulence transport by $\mathbf{E}\times\mathbf{B}$ rotation shear.Comment: accepted for publication in Nuclear Fusio

Comparison of Theory with Rotation Measurements in JET ICRH Plasmas

Plasma rotation appears to improve plasma performance by increasing the E x B flow shearing rate, thus decreasing radial correlations in the microturbulence. Also, plasma rotation can increase the stability to resistive MHD modes. In the Joint European Torus (JET), toroidal rotation rates omega (subscript ''tor'') with high Mach numbers are generally measured in NBI-heated plasmas (since the neutral beams aim in the co-plasma current direction). They are considerably lower with only ICRH (and Ohmic) heating, but still surprisingly large considering that ICRH appears to inject relatively small amounts of angular momentum. Either the applied torques are larger than naively expected, or the anomalous transport of angular momentum is smaller than expected. Since ICRH is one of the main candidates for heating next-step tokamaks, and for creating burning plasmas in future tokamak reactors, this paper attempts to understand ICRH-induced plasma rotation

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Current ramps in tokamaks: from present experiments to ITER scenarios

In order to prepare adequate current ramp-up and ramp-down scenarios for ITER, present experiments from various tokamaks have been analysed by means of integrated modelling in view of determining relevant heat transport models for these operation phases. A set of empirical heat transport models for L-mode (namely, the Bohm-gyroBohm model and scaling based models with a specific fixed radial shape and energy confinement time factors of H(96-L) = 0.6 or H(IPB98) = 0.4) has been validated on a multi-machine experimental dataset for predicting the l(i) dynamics within +/- 0.15 accuracy during current ramp-up and ramp-down phases. Simulations using the Coppi-Tang or GLF23 models (applied up to the LCFS) overestimate or underestimate the internal inductance beyond this accuracy (more than +/- 0.2 discrepancy in some cases). The most accurate heat transport models are then applied to projections to ITER current ramp-up, focusing on the baseline inductive scenario (main heating plateau current of I(p) = 15 MA). These projections include a sensitivity study to various assumptions of the simulation. While the heat transport model is at the heart of such simulations (because of the intrinsic dependence of the plasma resistivity on electron temperature, among other parameters), more comprehensive simulations are required to test all operational aspects of the current ramp-up and ramp-down phases of ITER scenarios. Recent examples of such simulations, involving coupled core transport codes, free-boundary equilibrium solvers and a poloidal field (PF) systems controller are also described, focusing on ITER current ramp-down.</p