Resonance Heating Research Articles

Assessment of plasma power deposition on the ITER ICRH antennas

Ion cyclotron resonance heating (ICRH) is one of the three additional heating schemes to be deployed on ITER. Its two antenna arrays, installed on the outboard midplane, will deliver 20 MW of RF power in the 40–55 MHz frequency range. The plasma-facing component of the antenna assembly is the Faraday screen, comprising beryllium (Be) tile armoured, actively cooled bars located only ~1 cm radially behind the innermost point of the shaped Be first wall panels (FWPs). As such they are in close proximity to the scrape-off layer (SOL) plasma and it is important to assess the maximum heat loads that the screen bars may experience during high power ITER operation. This paper provides a detailed assessment of these loads using the new 3D field line tracing and power deposition framework SMITER (Kos et al., 2019). The focus is on the H-mode, burning plasma scenario, taking into account both plasma heat loading (including average loading due to mitigated Type I ELMs) and the loads due to photonic impact (assessed with the optical ray-tracing package Raysect (Meakins and Carr, 2017)) from power radiated in the core obtained from integrated JINTRAC simulations. Calculations are also performed to assess the minimum allowed antenna to magnetic separatrix distances, for cases in which closer approach may be required to improve RF coupling.

Trapped upper hybrid waves as eigenmodes of non-monotonic background density profiles

Non-monotonic plasma density structures such as blobs and magnetic islands give rise to trapped upper hybrid (UH) waves. Trapped UH waves which satisfy Bohr–Sommerfeld quantization can be thought of as eigenmodes of a cavity. Using fully kinetic particle-in-cell simulations, we verify the existence of these UH eigenmodes and demonstrate their significance as only eigenfrequencies become unstable to three-wave interactions. The eigenmodes can be excited through parametric decay instabilities (PDIs) of an X-mode pump wave at approximately twice the UH frequency, as could be the case for a gyrotron beam traversing a blob in a magnetically confined fusion plasma. We derive a closed expression for the wavenumber of UH waves, which is accurate both close to the UH layer and to the electron cyclotron resonance. This allows for fast analysis of eigenmodes in a non-monotonic structure. An expression for the amplification of PDI daughter waves in an inhomogeneous plasma is extended to a decay region where the first several derivatives vanish. From the amplification in a convective PDI, we estimate the growth rate of the absolute PDI involving the trapped waves. We show that the excitation of eigenmodes through PDIs in our simulations are indeed absolute rather than convective due to the trapping of the daughter waves. Additionally, we show that only eigenmodes get excited through the PDIs, and that we are able to predict the growth rates of the daughter waves and how they scale with the pump wave intensity. This is evidence supporting a fundamental assumption of analytical theory describing low threshold strong scattering observed in magnetically confined fusion experiments during second harmonic electron cyclotron resonance heating (ECRH). Such low threshold instabilities can degrade ECRH performance but also offer novel uses for ion heating or as diagnostics.

Stability analysis of TJ-II stellarator NBI driven Alfvén eigenmodes in ECRH and ECCD experiments

In this paper, we analyze the impact of electron cyclotron resonance heating and electron cyclotron current drive on the Alfvénic instabilities driven by neutral beam injection observed in the TJ-II stellarator. An MHD stability analysis of driven Alfvén eigenmodes compatible with the experimental plasma parameters is carried out in order to compare with the data provided by magnetic coils, radiation monitors, and heavy ion beam probes. To this end, the vacuum magnetic configuration modified by the different levels of plasma current, the thermal plasma parameters and the fast ion pressure profiles generated by the co-injected neutral beam, are entered in the FAR3d gyro-fluid code in order to follow the linear evolution of the destabilized plasma equilibrium. Linear growth rates and radial location of the dominant predicted modes coincident in frequency with the observed fluctuations are presented. Despite the uncertainties related to the estimation of the rotational transform profile, the code predictions agree within reasonable accuracy with the experimental results.

First spatially resolved measurements of the D-3He α-particle source with the upgraded JET gamma-ray camera.

The Joint European Torus (JET) gamma-ray camera has been recently upgraded with the installation of new gamma-ray detectors, based on LaBr3(Ce) scintillation crystals, which add spectroscopic capability to the existing system allowing measurements with good energy resolution (5% at 0.622MeV), a dynamic range from hundreds of keV up to about 30MeV, and high counting rate capabilities of MCps. First gamma-ray measurements during the C38 campaign of the JET have been successfully carried out, in particular, in D-3He plasmas from three-ion ion cyclotron resonance heating experiments, where the detection of 16.4MeV γ-rays from D + 3He → γ + 5Li reactions with the gamma-ray camera upgrade allowed determining the spatial profile of alpha particles born in D + 3He fusion reactions.

Energy-angle distribution of the ions in the RF sheath of ICRH antennas

Radio-frequency sheaths forming at the surface of ion cyclotron resonance heating (ICRH) actuators in fusion experiments are associated with enhanced impurity sputtering from ICRH plasma-facing components (PFCs). The minimization of impurity fluxes from the ICRH PFC is a vital task for the usage of ICRH systems. Capturing the ion kinetics is important for an accurate description of plasma–material interactions, because the ion dynamics plays a crucial role in RF sheaths. Here, we present a hybrid particle-in-cell (hPIC) model able to capture the kinetic behavior of the ions. We analyze the kinetic ion energy-angle distributions (IEADs) impacting the RF antenna and its dependence on different plasma and RF sheath parameters. In particular, the IEAD dependence on RF frequency and magnetic field alignment is investigated. Using hPIC, we simulated a case emulating the latest experimental campaign from JET. The simulation showed that under specific plasma and RF parameters, the kinetic motion of ions results in a cusp formation in the IEAD.

Energy gain of beam-plasma D–T reaction in the presence of ICRH

A model for studying parametric dependence of the local energy gain (the ratio of fusion power density to absorbed power density) of a beam-plasma fusion reaction is proposed. It assumes that beam ions are produced by Neutral Beam Injection (NBI) and accelerated by Ion Cyclotron Resonance Heating (ICRH). The model includes several relations (for the local fusion energy gain, fusion power, and ICRH power absorbed) which employ a beam energy distribution function found analytically. The influence of ICRH on the beam ions is described by a quasilinear theory. Specific calculations are carried out for a deuterium beam with the ions born at energy Eb=100 keV and a tritium plasma. It is found that only those scenarios for which ICRH accelerates mainly slightly thermalized injected ions can increase the beam-plasma energy gain during NBI + ICRH significantly. In contrast, when the minimum energy of the ions affected by ICRH is much lower than the birth energy, ICRH does not enlarge the energy gain. On the other hand, fusion power increases due to ICRH, independently of the heating scheme. It is shown that the NBI energy structure is a factor which requires the maximum energy of injected ions in a certain range (150–200 keV when deuterium is injected into in a tritium plasma) for the beam-plasma energy gain to exceed unity significantly in both NBI and NBI + ICRH cases.

Comparison between fluctuation of floating potential gradient and velocity of blob structure on HL-2A tokamak

Several results based on the Langmuir probes’ data on the HL-2A tokamak are presented. The blob structures’ radial and poloidal drift velocities, estimated by the gradient of floating potential and by time delay evaluation, are compared in different line-averaged density and electron cyclotron resonance heating conditions. A positive correlation is observed in the comparison between blobs’ radial velocity estimated by the two methods mentioned above, regardless of the situation differences mentioned above. Correlation is also observed in the comparison between the blobs’ poloidal velocity estimated by the two methods in different situations, while a shift due to the different line-averaged density is observed. These results imply that the radial gradient of floating potential may have some value as a reference during data analysis in low-parameter discharge.

Thermographic reconstruction of heat load on the first wall of Wendelstein 7-X due to ECRH shine-through power

Electron cyclotron resonance heating (ECRH) is a powerful and flexible plasma heating technique that serves as the main heater at Wendelstein 7-X (W7-X) and will be used at ITER for start-up, heating, current drive and mitigation of plasma instabilities. In the case of poor or degraded microwave absorption, which is expected in the O2-mode heating scenario, a significant part of the beam directly hits the wall, leading to local overheating and potential damage. The ECRH shine-through power is mostly reflected onto the targets; only a small fraction is really absorbed through ohmic losses (typically 3% for graphite at 140 GHz). The ohmic losses do not only depend on the material properties and the frequency, but also on the polarization of the wave and the angle of incidence. This paper presents a thermographic analysis of ECRH experiments at W7-X, including heat load and temperature simulations of the first wall that include ECRH shine through. Two O-mode ECRH experiments with both a high temperature rise of the first wall and different angles of beam incidence on the wall’s surface are depicted. One experiment has 775 kW of power modulation (5 Hz) with mixed polarization (45% O-mode, 55% X-mode) and an EC beam angle almost normal to the first wall. The second has 550 kW of steady EC power with O-mode polarization, a shallow beam angle and increased power absorption by the material. It is shown that infrared thermography is a useful tool for measuring shine-through power and protecting wall components.

Generation of 1.5 MW–140 GHz Pulses With the Modular Pre-Prototype Gyrotron for W7-X

In anticipation of an Electron Cyclotron Resonance Heating system upgrade for the stellarator Wendelstein 7-X, a 1.5 MW - 140 GHz continuous-wave gyrotron is under development. In order to provide a first experimental verification of the scientific RF and electron beam optics design of the gyrotron with ms pulses, the Karlsruhe Intitule of Technology has developed a short-pulse pre-prototype gyrotron. In this work, we present details regarding the construction of the pre-prototype as well as measurements from the first experimental campaign delivering up to 1.6 MW in short pulses.

Mechanical and electromagnetic design of the vacuum vessel of the SMART tokamak

The SMall Aspect Ratio Tokamak (SMART) is a new spherical device that is currently being designed at the University of Seville. SMART is a compact machine with a plasma major radius (R) greater than 0.4 m, plasma minor radius (a) greater than 0.2 m, an aspect ratio (A) over than 1.7 and an elongation (k) of more than 2. It will be equipped with 4 poloidal field coils, 4 divertor field coils, 12 toroidal field coils and a central solenoid. The heating system comprises of a Neutral Beam Injector (NBI) of 600 kW and an Electron Cyclotron Resonance Heating (ECRH) of 6 kW for pre-ionization. SMART has been designed for a plasma current (Ip) of 500 kA, a toroidal magnetic field (BT) of 1 T and a pulse length of 500 ms preserving the compactness of the machine. The free boundary equilibrium solver code FIESTA [1] coupled to the linear time independent, rigid plasma model RZIP [2] has been used to calculate the target equilibria taking into account the physics goals, the required plasma parameters, vacuum vessel structures and power supply requirements. We present here the final design of the SMART vacuum vessel together with the Finite Element Model (FEM) analysis carried out to ensure that the tokamak vessel provides high quality vacuum and plasma performance withstanding the electromagnetic j×B loads caused by the interaction between the eddy currents induced in the vessel itself and the surrounding magnetic fields. A parametric model has been set up for the topological optimization of the vessel where the thickness of the wall has been locally adapted to the expected forces. An overview of the new machine is presented here.

Recent progress on neoclassical impurity transport in stellarators with implications for a stellarator reactor

Accumulation of impurities in the core of the plasma is a potential problem for any fusion reactor, and is thought to be especially problematic for stellarators. We use the collisional transport code Sfincs to calculate, in a reactor stellarator scenario, the effects of flux-surface electrostatic potential variation on tungsten transport. The potential variation is found to have a large effect on the transport, reversing the sign of the transport coefficients related to the radial electric field and the bulk ion temperature gradient, and also increasing the magnitude of the former by two orders of magnitude. We optimize the electrostatic potential variation to minimize impurity accumulation, resulting in a 20%–50% reduction of the inward impurity flux. Furthermore, we investigate whether the electrostatic potential variation caused by fast particles due to ion-cyclotron resonance heating can be used to reduce the impurity accumulation, but find the effect to be small in the cases considered.

Preliminary RF characterization of DEMO in-port ion cyclotron heating system for toroidal and poloidal geometry variations

Ion cyclotron Resonance Heating (ICRH) is one of the auxiliary heating schemes considered for the future DEMOnstration Power Plant (DEMO). The ICRH system in DEMO should couple 50 MW using three equatorial ports. Compared to the ITER one, the current DEMO port design has a smaller toroidal and a larger poloidal extent.One of the ICRH pre-conceptual design options for DEMO, an in-port antenna based on the current ITER design, is investigated using the ANTITER II code. To make optimal use of the port larger poloidal extent, the possibility to use quadruplets instead of the ITER triplets is explored. The impact on the power coupling is evaluated for two different antenna geometry considering a possible toroidal extension of the antenna in the blanket. The ITER matching-decoupling system is adapted to the new antenna front geometries considered.

Design and thermal-structural analysis of a high power ICRH travelling wave array antennas

Travelling wave array (TWA) antennas have been proposed for the ICRH (Ion Cyclotron Resonance Heating) antennas of fusion reactors in view to decreasing the antenna voltage and associated electric field. This paper reports the progress of the design and structural analysis of an actively cooled high power TWA antenna for WEST. First, the design of a non-cooled mock-up that will be tested in the TITAN facility is presented. The main objective is to assess the voltage stand-off of the antenna at power and electric field levels relevant for future nominal operation in a fusion device. The main characteristics of the mock-up are detailed and the results from thermal and structural analysis of the mock-up tested in TITAN are presented. In the second part, the compatibility with WEST environment for an actively-cooled TWA antenna is assessed. Mechanical, electromagnetic, thermal and hydraulics constrains (.i.e VDE, plasma radiation, toroidal magnetic field ripple…) are listed in a system loads specification. A Virtual Reality analysis using IRFM tools has been performed to check compliance of the design in term of assembly inside the WEST vacuum vessel.

First results from the NORTH tokamak

NORTH is a small-sized tokamak located at the Technical University of Denmark. It can be operated both as a spherical tokamak and a simple magnetized torus (SMT). Here we show first experimental results. We discuss the setup of the device, including the layout of the vacuum vessel, the magnetic field coils, the power supply systems, the heating sources, diagnostics, and the control system. We present measurements in discharges that have been heated by electron cyclotron resonance heating at the fundamental resonance. The waves have been injected both from the low-field side and the high-field side, and rich fluctuation spectra in Langmuir probe and Rogowski coil signals are presented during the different plasma phases. Finally, we discuss research plans and further installations to be made on NORTH.

Neutron and gamma-ray transport calculations in support of the design of the radiation shielding for the TOFED neutron spectrometer at LHD

The neutron emission spectrometer (NES) has been commissioned at the Large Helical Device (LHD) with a new Time-Of-Flight-Enhanced-Diagnostics (TOFED). The aim of the TOFED is to measure the neutron energy spectrum to study the fast-ions behavior such as high-energy ions created by ion cyclotron resonance heating, beam ions, and D-D fusion born 1 MeV tritons in LHD. The instrument is assembled of the first detector S1 and the second detector S2. In 2019, TOFED was installed at Z = 13.5 m below midplane without the radiation shielding and was initially commissioned. Due to neutrons streaming through a considerable number of holes in the LHD torus hall, the ambient background radiation at the TOFED was higher than expected, which causes the large number of coincidental events on the detectors. The obtained time difference between S1 and S2 provides the strong suggestion of TOFED relocation and of TOFED shielding development. In 2020, TOFED was relocated to the basement at Z = 18.5 m below midplane. In this work, a dedicated radiation shielding design is investigated using the Monte Carlo N-Particle code (MCNP6). The shielding design fulfills the design criteria on the radiation moderation capability and engineering constraints. Due to the design criteria, borated polyethylene with thickness of 20 cm is being considered as a neutron shielding and lead with thickness of 5 cm is being considered as a gamma-ray shielding. The shielding is suggested to be a rectangular polyhedron shaped roof. With the designed shielding, neutron flux is reduced by approximately two orders of magnitude at S1 and by approximately more than two orders of magnitude at S2 while prompt gamma-ray fluence is reduced by approximately two orders of magnitude. Here, the detailed design of radiations shielding for the TOFED at LHD is presented. The appropriate radiation shielding from this work will be installed at TOFED.

A New Lumped Approach for the Simulation of the Magnetron Injection Gun for MegaWatt-Class EU Gyrotrons

In the framework of the ITER (International Thermonuclear Experimental Reactor) project, one of the key components of the reactor is the ECRH (Electron Cyclotron Resonance Heating). This system has the duty to heat the plasma inside the tokamak, using high frequency and power radio waves, produced by sets of 1MW gyrotrons. One of the main issues related to the gyrotron operation is the output power drop that happens right after the beginning of a pulse. In this work, we study the underlying phenomena that cause the power drop, focusing on the gyrotron’s MIG (Magnetron Injection Gun) of the 1MW, 170 GHz European Gyrotron prototype for ITER. It is shown how the current emission and the temperature of the emitter are tightly bound, and how their interaction causes the power drop, observed experimentally. Furthermore, a simple yet effective lumped-parameter model to describe the MIG’s cathode thermal dynamics is developed, which is able to predict the power output of the gyrotron by simulating the propagation of the heat inside this component. The model is validated against test results, showing a good capability to reproduce the measured behavior of the system, while still being open to further improvements.

Kalman filter density reconstruction in ICRH discharges on ASDEX Upgrade

Plasma density is one of the key quantities that need to be controlled in real-time as it scales directly with fusion power and, if left uncontrolled, density limits can be reached leading to a disruption. On ASDEX Upgrade (AUG), the real-time measurements are the line-integrated density, measured by the interferometers, and the average density derived from the bremsstrahlung measured by spectroscopy. For control, these measurements are used to reconstruct the radial density profile using an extended Kalman filter (EKF). However, in discharges where ion cyclotron resonance heating (ICRH) is used, the measurements from the interferometers are corrupted and the reconstructed density is false. In this paper, the existing EKF implementation is improved, implemented and experimentally verified on AUG. The new EKF includes a new particle transport model in the prediction model RAPDENS as well as a new representation of ionization and recombination. Furthermore, an algorithm was introduced that is capable of detecting the corrupt diagnostics; this algorithm is based on the rate of change of the innovation residual. The changes to the RAPDENS observer resulted in better density reconstruction in ICRH discharges where corrupt measurement occur. The new version has been implemented on the real-time control system at AUG and functions properly in ICRH discharges.

Tangential hard x-ray diagnostic array on the EXL-50 spherical tokamak.

A tangential hard x-ray (HXR) diagnostic on the newly constructed ENN XuanLong-50 (EXL-50) spherical tokamak for fast electron emission studies is presented. The HXR detection system consists of a symmetrical CdZnTe semiconductor detector array with a spectral sensitivity range of 20-300 keV. 25 channels have been designed on the 270° horizontal vacuum port with 12 sight lines to observe the forward emission, 12 sight lines to observe the backward emission of fast electrons, and 1 for viewing the central. Currently, ten channels have been in operation in the EXL-50 experiments. The systems are designed to measure the x-ray spectra for the estimation of fast electron temperature and electron velocity distribution in the EXL-50 experiment, which will be useful for understanding the dynamics of fast electrons generated by electron cyclotron resonance heating, for plasma instability and transport studies and for the analysis of plasma heating efficiency.

Heat pulse propagation and anomalous electron heat transport measurements on the optimized stellarator W7-X

The optimized stellarator Wendelstein 7-X (W7-X) is designed to have an approximately quasi-isodynamic magnetic configuration with reduced neoclassical transport in comparison to a classical stellarator, and turbulent transport is expected to be a significant source of anomalous heat transport across the plasma minor radius. The ion temperature gradient driven mode and the trapped electron mode (TEM) are thought to be responsible for the ion-scale turbulence in W7-X plasmas with volume averaged pressure below 1%. In this work, the electron temperature gradient driven turbulence is shown to be a good candidate for the explanation of the observed electron heat flux, in the inner plasma region where the density gradient is weak (in the outer region, a relatively stronger density gradient would drive additional TEM turbulence). The experimental electron heat transport measured during electron cyclotron resonant heating power and plasma density scans is compared to neoclassical predictions, and the stiffness in the electron heat transport measured during transient transport experiments is presented in three common magnetic configurations of W7-X. In low-⟨β⟩ plasma discharges, the stiffness in the electron heat flux, quantified by the ratio of the heat pulse to power balance diffusivity, , is measured to be less than 2, and trend downwards with increasing collisionality.

JET ILA full array and polychromatic operation

The JET ILA is an Ion Cyclotron Resonance Heating (ICRH) antenna composed of four resonant double loops (RDLs) arranged in a 2 toroidal by 2 poloidal compact array. It has been operated in a range of frequencies from 29 to 51 MHz, but mostly on half array. Some full array pulses were achieved but without all control loops implemented.All control algorithms (first and second stage matching, toroidal and poloidal phasing, totalling 22 feedback loops) have now been implemented and recently allowed to operate the full array at 33 MHz and 42 MHz up to 4 MW. It also allowed for the first time to perform a poloidal phase scan of the antenna at 37 MHz, with dipole toroidal phasing.Although full array operation has mostly proven reliable at 33 MHz, it has proven unstable at 42 MHz in several occasions due to the high level of mutual coupling between the different RDLs inherent to a compact array. To decouple the control of the top and bottom halves, first tests were successfully performed with polychromatic – or dual frequency – operation of the ILA (operation at 2 different frequencies around a central frequency).