The TCV tokamak is augmenting its unique historical capabilities (strong shaping, strong electron heating) with ion heating, additional electron heating compatible with high densities, and variable divertor geometry, in a multifaceted upgrade program designed to broaden its operational range without sacrificing its fundamental flexibility. The TCV program is rooted in a three-pronged approach aimed at ITER support, explorations towards DEMO, and fundamental research. A 1 MW, tangential neutral beam injector (NBI) was recently installed and promptly extended the TCV parameter range, with record ion temperatures and toroidal rotation velocities and measurable neutral-beam current drive. ITER-relevant scenario development has received particular attention, with strategies aimed at maximizing performance through optimized discharge trajectories to avoid MHD instabilities, such as peeling-ballooning and neoclassical tearing modes. Experiments on exhaust physics have focused particularly on detachment, a necessary step to a DEMO reactor, in a comprehensive set of conventional and advanced divertor concepts. The specific theoretical prediction of an enhanced radiation region between the two X-points in the low-field-side snowflake-minus configuration was experimentally confirmed. Fundamental investigations of the power decay length in the scrape-off layer (SOL) are progressing rapidly, again in widely varying configurations and in both D and He plasmas; in particular, the double decay length in L-mode limited plasmas was found to be replaced by a single length at high SOL resistivity. Experiments on disruption mitigation by massive gas injection and electron-cyclotron resonance heating (ECRH) have begun in earnest, in parallel with studies of runaway electron generation and control, in both stable and disruptive conditions; a quiescent runaway beam carrying the entire electrical current appears to develop in some cases. Developments in plasma control have benefited from progress in individual controller design and have evolved steadily towards controller integration, mostly within an environment supervised by a tokamak profile control simulator. TCV has demonstrated effective wall conditioning with ECRH in He in support of the preparations for JT-60SA operation.
The Electron Cyclotron (EC) system for the ITER tokamak is designed to inject ≥20 MW RF power into the plasma for Heating and Current Drive (H&CD) applications. The EC system consists of up to 26 gyrotrons (between 1 to 2 MW each), the associated power supplies, 24 transmission lines and 5 launchers. The EC system has a diverse range of applications including central heating and current drive, current profile tailoring and control of plasma magneto-hydrodynamic (MHD) instabilities such as the sawtooth and neoclassical tearing modes (NTMs). This diverse range of applications requires the launchers to be capable of depositing the EC power across nearly the entire plasma cross section. This is achieved by two types of antennas: an equatorial port launcher (capable of injecting up to 20 MW from the plasma axis to mid-radius) and four upper port launchers providing access from inside of mid radius to near the plasma edge. The equatorial launcher design is optimized for central heating, current drive and profile tailoring, while the upper launcher should provide a very focused and peaked current density profile to control the plasma instabilities.The overall EC system has been modified during the past three years taking into account the issues identified in the ITER design review from 2007 and 2008 as well as integrating new technologies. This paper will review the principal objectives of the EC system, modifications made during the past two years and how the design is compliant with the principal objectives.
Abstract-In this paper, we present a detailed analysis of the iterative phase retrieval approach (IPRA) for determining the phase profile of the output microwave beam of a gyrotron from known intensity patterns emphasizing the field propagation techniques which are used to propagate the RF field of the microwave beam between known intensity planes. The propagation method, based on first Rayleigh-Sommerfeld diffraction integral (RSDI), is solved using fast Fourier transform (FFT) technique and zero padding. It is observed that the use of FFT and, therefore, the discretization of the RSDI propagation kernel introduce aberrations in the propagated field due to the superposition of the original field with its replicated versions. This problem is solved by approximations leading to the Huygens-Fresnel propagation method which further imposes the restrictions on the distances of propagation depending on the size of the transverse plane used to discretize the intensity pattern. This constraint of the distance of propagation causes problem in the iterative phase retrieval approach (IPRA) when more than two intensity planes are used. A method based on interpolation is proposed to overcome this restriction. IPRA is then further discussed to optimize several parameters, such as plane separation, plane dimension, mesh size, and measurement accuracies, which become more of an issue during the measurements of infrared intensity thermograms of the output microwave beam.Index Terms-Fourier transform, gyrotron, infrared intensity thermogram, millimeter wave, phase retrieval, zero padding.
In the reference ECH design, ITER requires a total of 20MW/CW power at 170GHz using gyrotrons with a unit power of 1MW. A higher power per unit (2MW/gyrotron) would result in a strong reduction of the cost of the whole ECRH system, and would also relax the room constraints on the launcher antenna design. The high power capability of coaxial cavity gyrotrons has been demonstrated with short pulse experiments at FZK. A collaborative effort between European research associations CRPP-EPFL, FZK and Thales Electron Devices (TED) has been launched by the European Fusion Development Agreement (EFDA) in 2003, aiming at the development of an industrial 170GHz/2MW/CW coaxial cavity gyrotron. The first prototype, although designed to be CW compatible is expected to reach 2MW/1s and has been delivered by end of 2006. It will be tested in Lausanne, where a specially dedicated test facility has been built. The test facility has been designed to be flexible enough, allowing the possible commissioning of tubes with different characteristics, as well the tests of the launcher antenna at full performances. Initial experiments are planned for the end of the third quarter 2007.
Abstract. Understanding the propagation of high power mm-wave in plasmas is of tremendous importance in the route to fusion considering their extensive use in magnetically confined fusion devices. Mm-beams, launched from the outside of the vessel must propagate through plasma edge-turbulence before reaching their target region. Until recently, the effect of edge-turbulence on the beam propagation was neglected, but it has been estimated for ITER that it could lead to significant differences in the time-averaged and instantaneous beam profiles, leading to a loss of efficiency in their use. In this paper, we present first direct experimental measurements of high power beam after propagation in simple magnetized toroidal plasmas in TCV.
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