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SIDON (SImulator of raDiO-frequency Networks) is an in-house developed Radio-Frequency (RF) network solver that has been implemented to cross-validate the design of WEST ICRF launchers and simulate their impedance matching algorithm while considering all mutual couplings and asymmetries. In this paper, the authors illustrate the theory of SIDON as well as results of its calculations. The authors have built time-varying plasma scenarios (a sequence of launchers front-faces L-mode and H-mode Z-matrices), where at each time step (1 millisecond here), SIDON solves the RF network. At the same time, when activated, the impedance matching algorithm controls the matching elements (vacuum capacitors) and thus their corresponding S-matrices. Typically a 1-second pulse requires around 10 seconds of computational time on a desktop computer. These tasks can be hardly handled by commercial RF software. This innovative work allows identifying strategies for the launchers future operation while insuring the limitations on the currents, voltages and electric fields, matching and Load-Resilience, as well as the required straps voltage amplitude/phase balance. In this paper, a particular attention is paid to the simulation of the launchers behavior when arcs appear at several locations of their circuits using SIDON calculator. This latter work shall confirm or identify strategies for the arc detection using various RF electrical signals. One shall note that the use of such solvers in not limited to ICRF launchers simulations but can be employed, in principle, to any linear or linearized RF problem.

Launching electromagnetic waves in the ion cyclotron range of frequencies (ICRF) is an efficient method of plasma heating, actively employed in most of fusion machines. ICRF has a number of important supplementary applications, including the generation of high-energy ions. In this paper, we discuss a new set of three-ion ICRF scenarios and the prospect of their use as a dedicated tool for fast ion generation in tokamaks and stellarators. A distinct feature of these scenarios is a strong absorption efficiency possible at very low concentrations of resonant minority ions (∼ 1% or even below). Such concentration levels are typical for impurities contaminating fusion plasmas. An alternative ICRF scenario for maximizing the efficiency of bulk D-T ion heating is suggested for JET and ITER tokamaks, which is based on three-ion ICRF heating of intrinsic Beryllium impurities.

The paper reports results of MCNP-5 calculations to assess the effect on the Tritium Breeding Ratio (TBR) when integrating a distributed Ion Cyclotron Range of Frequencies (ICRF) antenna in the blanket of DEMO fusion power reactor. The calculations consider different parameters such as the ICRF covering ratio and the type of breeding blanket including the Helium Cooled Pebble Bed (HCPB) and the Helium Cooled Lithium Lead (HCLL) concepts. For an antenna with a full toroidal circumference of 360°, located poloidally at 40° with a poloidal extension of 1 m, the reduction of the TBR is −0.349% for the HCPB blanket and −0.532% for the HCLL blanket. The distributed ICRF antenna is thus shown to have only a marginal effect on the TBR of the DEMO reactor.

RF sheaths rectification can be the reason for operational limits for Ion Cyclotron Range of Frequencies (ICRF) heating systems via impurity production or excessive heat loads. To simulate this process in realistic geometry, the Self-consistent Sheaths and Waves for Ion Cyclotron Heating (SSWICH) code is a minimal set of coupled equations that computes self-consistently wave propagation and DC plasma biasing. The present version of its wave propagation module only deals with the Slow Wave assumed to be the source of RF sheath oscillations. However the ICRF power coupling to the plasma is due to the fast wave (FW). This paper proposes to replace this one wave equation module by a full wave module in either 2D or 3D as a first step towards integrated modelling of RF sheaths and wave coupling. Since the FW is propagative in the main plasma, Perfectly Matched Layers (PMLs) adapted for plasmas were implemented at the inner side of the simulation domain to absorb outgoing waves and tested numerically with tilted BD in Cartesian geometry, by either rotating the cold magnetized plasma dielectric tensors in 2D or rotating the coordinate vector basis in 3D. The PML was further formulated in cylindrical coordinates to account for for the toroidal curvature of the plasma. Toroidal curvature itself does not seem to change much the coupling. A detailed 3D geometrical description of Tore Supra and ASDEX Upgrade (AUG) antennas was included in the coupling code. The full antenna structure was introduced, since its toroidal symmetry with respect to the septum plane is broken (FS bars, toroidal phasing, non-symmetrical structure). Reliable convergence has been obtained with the density profile up to the leading edge of antenna limiters. Parallel electric field maps have been obtained as an input for the present version of SSWICH.

The longstanding expertise of the Tore Supra team in long pulse heating and current drive with radiofrequency (RF) systems will now be exploited in the WEST device (tungsten-W Environment in Steady-state Tokamak) [1]. WEST will allow an integrated long pulse tokamak programme for testing W-divertor components at ITER-relevant heat flux (10-20 MW/m2), while treating crucial aspects for ITER-operation, such as avoidance of W-accumulation in long discharges, monitoring and control of heat fluxes on the metallic plasma facing components (PFCs) and coupling of RF waves in H-mode plasmas. Scenario modelling using the METIS-code shows that ITER-relevant heat fluxes are compatible with the sustainment of long pulse H-mode discharges, at high power (up to 15 MW / 30 s at IP = 0.8 MA) or high fluence (up to 10 MW / 1000 s at IP = 0.6 MA) [2], all based on RF heating and current drive using Ion Cyclotron Resonance Heating (ICRH) and Lower Hybrid Current Drive (LHCD). This paper gives a description of the ICRH and LH...

The paper reports on latest developments for the ITER Ion Cyclotron Heating and Current Drive system: imminent acceptance tests of a prototype power supply at full power; successful factory acceptance of candidate RF amplifier tubes which will be tested on dedicated facilities; further design integration and experimental validation of transmission line components under 6MW hour-long pulses. The antenna Faraday shield thermal design has been validated above requirements by cyclic high heat flux tests. R&D on ceramic brazing is under way for the RF vacuum windows. The antenna port plug RF design is stable but major evolution of the mechanical design is in preparation to achieve compliance with the load specification, warrant manufacturability and incorporate late interface change requests. The antenna power coupling capability predictions have been strengthened by showing that, if the plasma scrape-off layer turns out to be steep and the edge density low, the reference burning plasma can realistically be displaced to improve the coupling.

In the present paper, the perturbation of the launched power spectrum of the Lower Hybrid wave at the separatrix by electron density fluctuations in the scrape-off layer is investigated. Considering a slab geometry with magnetic field lines parallel to the toroidal direction, the full wave equation is solved using Comsol Multiphysics r for a fully active multi-junction like LH antenna made of two modules. When electron density fluctuations are incorporated in the dielectric tensor over a thin perturbed layer in front of the grill, it is shown that the power spectrum may be strongly modified from the antenna mouth to the plasma separatrix as the wave propagates. The diffraction effect leads to the appearance of multiple satellite lobes with randomly varying positions, a feature consistent with the recently developed model that has been applied successfully to high density discharges on the Tokamak Tore Supra corresponding to the large spectral gap regime [Decker J. et al. Phys. Plasma 21 (2014) 092504]. The perturbation is found to be maximum for the Fourier components of the fluctuating spectrum in the vicinity of the launched LH wavelength.

In tokamak fusion plasmas, coherent fluctuations in the form of blobs or filaments are routinely observed in the scrape-off layer. In this paper we develop an analytical formalism for the scattering of radio frequency waves by filaments which are cylindrical with their major axis aligned along the toroidal magnetic field lines. Since the magnitude of the ratio of the density inside the filaments to the background density is generally of order 1, the geometric optics approximation cannot be used to describe the scattering. A full-wave model is formulated which assumes that the plasma is cold and that the plasma in the cylindrical filament has uniform density. The background plasma, in which the filament is present, is also assumed to be cold and uniform. The theoretical framework applies to the scattering of any plasma wave.

The ITER-like Antenna (ILA) for JET [1] is a 2 toroidal by 2 poloidal array of Resonant Double Loops (RDL) featuring in-vessel matching capacitors feeding RF current straps in conjugate-T manner, a low impedance quarter-wave impedance transformer, a service stub allowing hydraulic actuator and water cooling services to reach the aforementioned capacitors and a 2nd stage phase-shifter-stub matching circuit allowing to correct/choose the conjugate-T working impedance. Toroidally adjacent RDLs are fed from a 3dB hybrid splitter. It has been operated at 33, 42 and 47MHz on plasma (2008-2009) while it presently estimated frequency range is from 29 to 49MHz. At the time of the design (2001-2004) as well as the experiments the circuit models of the ILA were quite basic. The ILA front face and strap array Topica model was relatively crude and failed to correctly represent the poloidal central septum, Faraday Screen attachment as well as the segmented antenna central septum limiter. The ILA matching capacitors, T-junction, Vacuum Transmission Line (VTL) and Service Stubs were represented by lumped circuit elements and simple transmission line models. The assessment of the ILA results carried out to decide on the repair of the ILA identified that achieving routine full array operation requiresmore » a better understanding of the RF circuit, a feedback control algorithm for the 2nd stage matching as well as tighter calibrations of RF measurements. The paper presents the progress in modelling of the ILA comprising a more detailed Topica model of the front face for various plasma Scrape Off Layer profiles, a comprehensive HFSS model of the matching capacitors including internal bellows and electrode cylinders, 3D-EM models of the VTL including vacuum ceramic window, Service stub, a transmission line model of the 2nd stage matching circuit and main transmission lines including the 3dB hybrid splitters. A time evolving simulation using the improved circuit model allowed to design and simulate the effectiveness of a feedback control algorithm for the 2nd stage matching and demonstrates the simultaneous matching and control of the 4 RDLs: 11 feedback loops control 21 actuators (8 capacitors, 4 phase shifters and 4 stubs for the 2nd stage matching, 4 main phase shifters controlling of the toroidal phasing and the electronically controlled phase between RF sources feeding top and bottom parts of the array and determines the poloidal phasing of the array which is solved explicitly at each time step) on (simulated) ELMy plasmas.« less

The physics of wave heating and current drive processes in the bulk hot plasma are generally well identified. However, details of the wave-plasma interaction in the cold plasma edge are still not fully understood. In this paper a spectroscopic technique allowing for measurements of the electric field driving wave-plasma interactions in the edge region for LH current drive and ICRH systems will be discussed. Experimental results obtained in Tore Supra near the LH C3 antenna and in the magnetized capacitively coupled RF sheath from an experiment designed to mimic the ICRF near-field are presented. In conclusion, future plans to implement a laser-based spectroscopic technique to acquire the high resolution measurements needed for model validation is discussed.

ITER is an international project to design and build an experimental fusion reactor based on the “tokamak” concept. ITER relies upon localized electron cyclotron current drive (ECCD) at the rational safety factor q=2 to suppress or stabilize the expected poloidal mode m=2, toroidal mode n=1 neoclassical tearing mode (NTM) islands. Such islands if unmitigated degrade energy confinement, lock to the resistive wall (stop rotating), cause loss of “H-mode” and induce disruption. The International Tokamak Physics Activity (ITPA) on MHD, Disruptions and Magnetic Control joint experiment group MDC-8 on Current Drive Prevention/Stabilization of Neoclassical Tearing Modes started in 2005, after which assessments were made for the requirements for ECCD needed in ITER, particularly that of rf power and alignment on q=2 [1]. Narrow well-aligned rf current parallel to and of order of one percent of the total plasma current is needed to replace the “missing” current in the island O-points and heal or preempt (avoid destabilization by applying ECCD on q=2 in absence of the mode) the island [2-4]. This paper updates the advances in ECCD stabilization on NTMs learned in DIII-D experiments and modeling during the last 5 to 10 years as applies to stabilization by localized ECCD of tearing modes in ITER. This includes the ECCD (inside the q=1 radius) stabilization of the NTM “seeding” instability known as sawteeth (m/n=1/1) [5]. Recent measurements in DIII-D show that the ITER-similar current profile is classically unstable, curvature stabilization must not be neglected, and the small island width stabilization effect from helical ion polarization currents is stronger than was previously thought [6]. The consequences of updated assumptions in ITER modeling of the minimum well-aligned ECCD power needed are all-in-all favorable (and well-within the ITER 24 gyrotron capability) when all effects are included. However, a “wild card” may be broadening of the localized ECCD by the presence of the island; various theories predict broadening could occur and there is experimental evidence for broadening in DIII-D. Wider than now expected ECCD in ITER would make alignment easier to do but weaken the stabilization and thus require more rf power. In addition to updated modeling for ITER, advances in the ITER-relevant DIII-D ECCD gyrotron launch mirror control system hardware and real-time plasma control system have been made [7] and there are plans for application in DIII-D ITER demonstration discharges.