The demand for a predictive tool to help in designing ion-cyclotron radio frequency (ICRF) antenna systems for today's fusion experiments has driven the development of codes such as ICANT, RANT3D, and the early development of TOPICA (TOrino Polytechnic Ion Cyclotron Antenna) code. This paper describes the substantive evolution of TOPICA formulation and implementation that presently allow it to handle the actual geometry of ICRF antennas (with curved, solid straps, a general-shape housing, Faraday screen, etc) as well as an accurate plasma description, accounting for density and temperature profiles and finite Larmor radius effects. The antenna is assumed to be housed in a recess-like enclosure. Both goals have been attained by formally separating the problem into two parts: the vacuum region around the antenna and the plasma region inside the toroidal chamber. Field continuity and boundary conditions allow formulating of a set of two coupled integral equations for the unknown equivalent (current) sources; then the equations are reduced to a linear system by a method of moments solution scheme employing 2D finite elements defined over a 3D non-planar surface triangular-cell mesh. In the vacuum region calculations are done in the spatial (configuration) domain, whereas in the plasma region a spectral (wavenumber) representation of fields and currents is adopted, thus permitting a description of the plasma by a surface impedance matrix. Owing to this approach, any plasma model can be used in principle, and at present the FELICE code has been employed. The natural outcomes of TOPICA are the induced currents on the conductors (antenna, housing, etc) and the electric field in front of the plasma, whence the antenna circuit parameters (impedance/scattering matrices), the radiated power and the fields (at locations other than the chamber aperture) are then obtained. An accurate model of the feeding coaxial lines is also included. The theoretical model and its TOPICA implementation have been fully validated against measured data both in vacuo and in plasma-facing conditions for real-life structures.

The performance on plasma of the antennas of the proposed ITER ICRF system is evaluated by means of the antenna 24 × 24 impedance matrix provided by the TOPICA code and confirmed and interpreted by the semi-analytical code ANTITER II (summarized in an appendix). From this analysis the following system characteristics can be derived: (1) a roughly constant power capability in the entire 40–55 MHz frequency band with the same maximum voltage in the eight feeding lines is obtained for all the considered heating and current drive phasings on account of the broadbanding effect of service stubs. (2) The power capability of the array significantly depends on the distance of the antenna to the separatrix, the density profile in the scrape-off layer (SOL) and on the strap current toroidal and poloidal phasings. The dependence on phasing is stronger for wider SOL. (3) To exceed a radiated power capability of 20 MW per antenna array in the upper part of the frequency band, with a separatrix–wall distance of 17 cm and a conservative short decay plasma edge density profile, the system voltage stand-off must be 45 kV and well chosen combinations of toroidal and poloidal phasing are needed. (4) On account of the plasma gyrotropy and of poloidal magnetic field, special care must be taken in choosing the optimal toroidal current drive and poloidal phasings. The ANTITER II analysis shows furthermore that important coaxial and surface mode excitation can only be expected in the monopole toroidal phasing, that strong wave reflection from a steep density profile significantly reduces the coupling even if the separatrix is closer to the antenna and that the part of the edge density profile having a density lower than the cut-off density pertaining to the considered phasing does not significantly contribute to the coupling.

Équipe 107 : Physique des plasmas chaudsInternational audienceDuring the 2011 experimental campaign, one of the three ion cyclotron resonance heating (ICRH) antennas in the Tore Supra tokamak was equipped with a new type of Faraday screen (FS). The new design aimed at minimizing the integrated parallel electric field over long field lines as well as increasing the heat exhaust capability of the actively cooled screen. It proved to be inefficient for attenuating the radio-frequency (RF)-sheaths on the screen itself on the contrary to the heat exhaust concept that allowed operation despite higher heat fluxes on the antenna. In parallel, a new approach has been proposed to model self-consistently RF sheaths: the SSWICH (Self-consistent Sheaths and Waves for IC Heating) code. Simulations results from SSWICH coupled with the TOPICA antenna code were able to reproduce the difference between the two FS designs and part of the spatial pattern of heat loads and Langmuir probe floating potential. The poloidal pattern is a reliable result that mainly depends on the electrical design of the antenna while the radial pattern is on the contrary highly sensitive to loosely constrained parameters such as perpendicular conductivity that generates a DC current circulation from the private region inside the antenna limiters to the free scrape off layer outside these limiters. Moreover, the cantilevered bars seem to be the element in the screen design that enhanced the plasma potential

We present an ultrafast neural network (NN) model, QLKNN, which predicts core tokamak transport heat and particle fluxes. QLKNN is a surrogate model based on a database of 300 million flux calculations of the quasilinear gyrokinetic transport model QuaLiKiz. The database covers a wide range of realistic tokamak core parameters. Physical features such as the existence of a critical gradient for the onset of turbulent transport were integrated into the neural network training methodology. We have coupled QLKNN to the tokamak modelling framework JINTRAC and rapid control-oriented tokamak transport solver RAPTOR. The coupled frameworks are demonstrated and validated through application to three JET shots covering a representative spread of H-mode operating space, predicting turbulent transport of energy and particles in the plasma core. JINTRAC-QLKNN and RAPTOR-QLKNN are able to accurately reproduce JINTRAC-QuaLiKiz T i,e and n e profiles, but 3 to 5 orders of magnitude faster. Simulations which take hours are reduced down to only a few tens of seconds. The discrepancy in the final source-driven predicted profiles between QLKNN and QuaLiKiz is on the order 1%-15%. Also the dynamic behaviour was well captured by QLKNN, with differences of only 4%-10% compared to JINTRAC-QuaLiKiz observed at mid-radius, for a study of density buildup following the L-H transition. Deployment of neural network surrogate models in multi-physics integrated tokamak modelling is a promising route towards enabling accurate and fast tokamak scenario optimization, Uncertainty Quantification, and control applications.

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The JET 2019-2020 scientific and technological programme exploited the results of years of concerted scientific and engineering work, including the ITER-like wall (ILW: Be wall and W divertor) installed in 2010, improved diagnostic capabilities now fully available, a major Neutral Beam Injection (NBI) upgrade providing record power in 2019-2020, and tested the technical & procedural preparation for safe operation with tritium. Research along three complementary axes yielded a wealth of new results. Firstly, the JET plasma programme delivered scenarios suitable for high fusion power and alpha particle physics in the coming D-T campaign (DTE2), with record sustained neutron rates, as well as plasmas for clarifying the impact of isotope mass on plasma core, edge and plasma-wall interactions, and for ITER pre-fusion power operation. The efficacy of the newly installed Shattered Pellet Injector for mitigating disruption forces and runaway electrons was demonstrated. Secondly, research on the consequences of long-term exposure to JET-ILW plasma was completed, with emphasis on wall damage and fuel retention, and with analyses of wall materials and dust particles that will help validate assumptions and codes for design & operation of ITER and DEMO. Thirdly, the nuclear technology programme aiming to deliver maximum technological return from operations in D, T and D-T benefited from the highest D-D neutron yield in years, securing results for validating radiation transport and activation codes, and nuclear data for ITER.

The ITER Ion Cyclotron Heating and Current Drive system will deliver 20MW of radio frequency power to the plasma in quasi continuous operation during the different phases of the experimental programme. The system also has to perform conditioning of the tokamak first wall at low power between main plasma discharges. This broad range of reqiurements imposes a high flexibility and a high availabiUty. The paper highlights the physics and design reqiurements on the IC system, the main features of its subsystems, the predicted performance, and the current procurement and installation schedide.

A comparison of the ASDEX Upgrade 3-strap ICRF antenna data with the linear electromagnetic TOPICA calculations is presented. The comparison substantiates a reduction of the local electric field at the radially protruding plasma-facing elements of the antenna as a relevant approach for minimizing tungsten (W) sputtering in conditions when the slow wave is strongly evanescent. The measured reaction of the time-averaged RF current at the antenna limiters to the antenna feeding variations is less sensitive than predicted by the calculations. This is likely to have been caused by temporal and spatial fluctuations in the 3D plasma density distribution affected by local non-linear interactions. The 3-strap antenna with the W-coated limiters produces drastically less W sputtering compared to the W-coated 2-strap antennas. This is consistent with the non-linear asymptotic SSWICH-SW calculations for RF sheaths.