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

Formation of convective cells in the scrape-off layer of the CASTOR tokamak

Understanding of the scrape-off layer (SOL) physics in tokamaks requires diagnostics with sufficient temporal and spatial resolution. This contribution describes results of experiments performed in the SOL of the CASTOR tokamak (R=40 cm, a = 6 cm) by means of a ring of 124 Langmuir probes surrounding the whole poloidal cross section. The individual probes measure either the ion saturation current of the floating potential with the spatial resolution up to 3 mm. Experiments are performed in a particular magnetic configuration, characterized by a long parallel connection length in the SOL, L_par ~q2piR. We report on measurements in discharges, where the edge electric field is modified by inserting a biased electrode into the edge plasma. In particular, a complex picture is observed, if the biased electrode is located inside the SOL. The poloidal distribution of the floating potential appears to be strongly non-uniform at biasing. The peaks of potential are observed at particular poloidal angles. This is interpreted as formation of a biased flux tube, which emanates from the electrode along the magnetic field lines and snakes q times around the torus. The resulting electric field in the SOL is 2-dimensional, having the radial as well as the poloidal component. It is demonstrated that the poloidal electric field E_pol convects the edge plasma radially due to the E_pol x B_T drift either inward or outward depending on its sign. The convective particle flux is by two orders of magnitude larger than the fluctuation-induced one and consequently dominates.Comment: 12th International Congress on Plasma Physics, 25-29 October 2004, Nice (France

ELM triggering conditions for the integrated modeling of H-mode plasmas

Recent advances in the integrated modeling of ELMy H-mode plasmas are presented. A model for the H-mode pedestal and for the triggering of ELMs predicts the height, width, and shape of the H-mode pedestal and the frequency and width of ELMs. Formation of the pedestal and the L-H transition is the direct result of ExB flow shear suppression of anomalous transport. The periodic ELM crashes are triggered by either the ballooning or peeling MHD instabilities. The BALOO, DCON, and ELITE ideal MHD stability codes are used to derive a new parametric expression for the peeling-ballooning threshold. The new dependence for the peeling-ballooning threshold is implemented in the ASTRA transport code. Results of integrated modeling of DIII-D like discharges are presented and compared with experimental observations. The results from the ideal MHD stability codes are compared with results from the resistive MHD stability code NIMROD.Comment: 12th International Congress on Plasma Physics, 25-29 October 2004, Nice (France

Answers and Solutions

Answers and solutions to the puzzles in this issue

Recent EUROfusion Achievements in Support of Computationally Demanding Multiscale Fusion Physics Simulations and Integrated Modeling

Integrated modeling (IM) of present experiments and future tokamak reactors requires the provision of computational resources and numerical tools capable of simulating multiscale spatial phenomena as well as fast transient events and relatively slow plasma evolution within a reasonably short computational time. Recent progress in the implementation of the new computational resources for fusion applications in Europe based on modern supercomputer technologies (supercomputer MARCONI-FUSION), in the optimization and speedup of the EU fusion-related first-principle codes, and in the development of a basis for physics codes/modules integration into a centrally maintained suite of IM tools achieved within the EUROfusion Consortium is presented. Physics phenomena that can now be reasonably modelled in various areas (core turbulence and magnetic reconnection, edge and scrape-off layer physics, radio-frequency heating and current drive, magnetohydrodynamic model, reflectometry simulations) following successful code optimizations and parallelization are briefly described. Development activities in support to IM are summarized. They include support to (1) the local deployment of the IM infrastructure and access to experimental data at various host sites, (2) the management of releases for sophisticated IM workflows involving a large number of components, and (3) the performance optimization of complex IM workflows.This work has been carried out within the framework of the EUROfusion Consortium and has received funding from the Euratom research and training programme 2014 to 2018 under grant agreement 633053. The views and opinions expressed herein do not necessarily reflect those of the European Commission or ITER.Peer ReviewedPostprint (published version

Development and Validation of a Tokamak Skin Effect Transformer model

A control oriented, lumped parameter model for the tokamak transformer including the slow flux penetration in the plasma (skin effect transformer model) is presented. The model does not require detailed or explicit information about plasma profiles or geometry. Instead, this information is lumped in system variables, parameters and inputs. The model has an exact mathematical structure built from energy and flux conservation theorems, predicting the evolution and non linear interaction of the plasma current and internal inductance as functions of the primary coil currents, plasma resistance, non-inductive current drive and the loop voltage at a specific location inside the plasma (equilibrium loop voltage). Loop voltage profile in the plasma is substituted by a three-point discretization, and ordinary differential equations are used to predict the equilibrium loop voltage as function of the boundary and resistive loop voltages. This provides a model for equilibrium loop voltage evolution, which is reminiscent of the skin effect. The order and parameters of this differential equation are determined empirically using system identification techniques. Fast plasma current modulation experiments with Random Binary Signals (RBS) have been conducted in the TCV tokamak to generate the required data for the analysis. Plasma current was modulated in Ohmic conditions between 200kA and 300kA with 30ms rise time, several times faster than its time constant L/R\approx200ms. The model explains the most salient features of the plasma current transients without requiring detailed or explicit information about resistivity profiles. This proves that lumped parameter modeling approach can be used to predict the time evolution of bulk plasma properties such as plasma inductance or current with reasonable accuracy; at least in Ohmic conditions without external heating and current drive sources

Exploration of the equilibrium operating space for NSTX-Upgrade

This paper explores a range of high-performance equilibrium scenarios available in the NSTX-Upgrade device [J.E. Menard, submitted for publication to Nuclear Fusion]. NSTX-Upgrade is a substantial upgrade to the existing NSTX device [M. Ono, et al., Nuclear Fusion 40, 557 (2000)], with significantly higher toroidal field and solenoid capabilities, and three additional neutral beam sources with significantly larger current drive efficiency. Equilibria are computed with freeboundary TRANSP, allowing a self consistent calculation of the non-inductive current drive sources, the plasma equilibrium, and poloidal field coil current, using the realistic device geometry. The thermal profiles are taken from a variety of existing NSTX discharges, and different assumptions for the thermal confinement scalings are utilized. The no-wall and idealwall n=1 stability limits are computed with the DCON code. The central and minimum safety factors are quite sensitive to many parameters: they generally increases with large outer plasmawall gaps and higher density, but can have either trend with the confinement enhancement factor. In scenarios with strong central beam current drive, the inclusion of non-classical fast ion diffusion raises qmin, decreases the pressure peaking, and generally improves the global stability, at the expense of a reduction in the non-inductive current drive fraction; cases with less beam current drive are largely insensitive to additional fast ion diffusion. The non-inductive current level is quite sensitive to the underlying confinement and profile assumptions. For instance, for BT=1.0 T and Pinj=12.6 MW, the non-inductive current level varies from 875 kA with ITER-98y,2 thermal confinement scaling and narrow thermal profiles to 1325 kA for an ST specific scaling expression and broad profiles. This sensitivity should facilitate the determination of the correct scaling of transport with current and field to use for future fully non-inductive ST devices. Scenarios are presented which can be sustained for 8-10 seconds, or (20-30)τCR, at βN=3.8-4.5, facilitating, for instance, the study of disruption avoidance for very long pulse. Scenarios have been documented which can operate with βT~25% and equilibrated qmin>1. The value of qmin can be controlled at either fixed non-inductive fraction of 100% or fixed plasma current, by varying which beam sources are used, opening the possibility for feedback qmin control. In terms of quantities like collisionality, neutron emission, non-inductive fraction, or stored energy, these scenarios represent a significant performance extension compared to NSTX and other present spherical torii

Integrated core–SOL–divertor modelling for ITER including impurity: effect of tungsten on fusion performance in H-mode and hybrid scenario

The compatibility of two operational constraints—operation above the L–H power threshold and at low power to divertor—is examined for ITER long pulse H-mode and hybrid scenarios in integrated core–scrape off layer (SOL)–divertor modelling including impurities (intrinsic Be, He, W and seeded Ne). The core thermal, particle and momentum transport is simulated with the GLF23 transport model tested in the self-consistent simulations of temperatures, density and toroidal rotation velocity in JET hybrid discharges and extrapolated to ITER. The beneficial effect of toroidal rotation velocity on fusion gain is shown. The sensitivity studies with respect to operational (separatrix and pedestal density, Ne gas puff) and unknown physics (W convective velocity and perpendicular diffusion in SOL as well as W prompt re-deposition) parameters are performed to determine their influence on the operational window and fusion gain.</p