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.

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.

FAST is a new machine proposed to support ITER experimental exploitation as well as to anticipate DEMO relevant physics and technology. FAST is aimed at studying, under burning plasma relevant conditions, fast particle (FP) physics, plasma operations and plasma wall interaction in an integrated way. FAST has the capability to approach all the ITER scenarios significantly closer than the present day experiments using deuterium plasmas. The necessity of achieving ITER relevant performance with a moderate cost has led to conceiving a compact tokamak (R = 1.82 m, a = 0.64 m) with high toroidal field (B T up to 8.5 T) and plasma current (I p up to 8 MA). In order to study FP behaviours under conditions similar to those of ITER, the project has been provided with a dominant ion cyclotron resonance heating system (ICRH; 30 MW on the plasma). Moreover, the experiment foresees the use of 6 MW of lower hybrid (LHCD), essentially for plasma control and for non-inductive current drive, and of electron cyclotron resonance heating (ECRH, 4 MW) for localized electron heating and plasma control. The ports have been designed to accommodate up to 10 MW of negative neutral beams (NNBI) in the energy range 0.5-1 MeV. The total power input will be in the 30-40 MW range under different plasma scenarios with a wall power load comparable to that of ITER (P /R ∼ 22 MW m −1). All the ITER scenarios will be studied: from the reference H mode, with plasma edge and ELMs characteristics similar to the ITER ones (Q up to ≈1.5), to a full current drive scenario, lasting around 170 s. The first wall (FW) as well as the divertor plates will be of tungsten in order to ensure reactor relevant

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.