Ion Cyclotron Resonance Heating Research Articles

Real-time diamagnetic flux measurements on ASDEX Upgrade.

Real-time diamagnetic flux measurements are now available on ASDEX Upgrade. In contrast to the majority of diamagnetic flux measurements on other tokamaks, no analog summation of signals is necessary for measuring the change in toroidal flux or for removing contributions arising from unwanted coupling to the plasma and poloidal field coil currents. To achieve the highest possible sensitivity, the diamagnetic measurement and compensation coil integrators are triggered shortly before plasma initiation when the toroidal field coil current is close to its maximum. In this way, the integration time can be chosen to measure only the small changes in flux due to the presence of plasma. Two identical plasma discharges with positive and negative magnetic field have shown that the alignment error with respect to the plasma current is negligible. The measured diamagnetic flux is compared to that predicted by TRANSP simulations. The poloidal beta inferred from the diamagnetic flux measurement is compared to the values calculated from magnetic equilibrium reconstruction codes. The diamagnetic flux measurement and TRANSP simulation can be used together to estimate the coupled power in discharges with dominant ion cyclotron resonance heating.

Density limits investigation and high density operation in EAST tokamak

Increasing the density in a tokamak is limited by the so-called density limit, which is generally performed as an appearance of disruption causing loss of plasma confinement, or a degradation of high confinement mode which could further lead to a H → L transition. The L-mode and H-mode density limit has been investigated in EAST tokamak. Experimental results suggest that density limits could be triggered by either edge cooling or excessive central radiation. The L-mode density limit disruption is generally triggered by edge cooling, which leads to the current profile shrinkage and then destabilizes a 2/1 tearing mode, ultimately resulting in a disruption. The L-mode density limit scaling agrees well with the Greenwald limit in EAST. The observed H-mode density limit in EAST is an operational-space limit with a value of . High density H-mode heated by neutral beam injection (NBI) and lower hybrid current drive (LHCD) are analyzed, respectively. The constancy of the edge density gradients in H-mode indicates a critical limit caused perhaps by e.g. ballooning induced transport. The maximum density is accessed at the H → L transition which is generally caused by the excessive core radiation due to high Z impurities (Fe, Cu). Operating at a high density () is favorable for suppressing the beam shine through NBI. High density H-mode up to could be sustained by 2 MW 4.6 GHz LHCD alone, and its current drive efficiency is studied. Statistics show that good control of impurities and recycling facilitate high density operation. With careful control of these factors, high density up to 0.93 stable H-mode operation was carried out heated by 1.7 MW LHCD and 1.9 MW ion cyclotron resonance heating with supersonic molecular beam injection fueling.

Observations of compound sawteeth in ion cyclotron resonant heating plasma using ECE imaging on experimental advanced superconducting tokamak

The spatial and temporal evolutions of compound sawteeth were directly observed using 2D electron cyclotron emission imaging on experimental advanced superconducting tokamak. The compound sawtooth consists of partial and full collapses. After partial collapse, the hot core survives as only a small amount of heat disperses outwards, whereas in the following full collapse a large amount of heat is released and the hot core dissipates. The presence of two q = 1 surfaces was not observed. Instead, the compound sawtooth occurs mainly at the beginning of an ion cyclotron resonant frequency heating pulse and during the L-H transition phase, which may be related to heat transport suppression caused by a decrease in electron heat diffusivity.

Estimates of RF-induced erosion at antenna-connected beryllium plasma-facing components in JET

Radio-frequency (RF)-enhanced surface erosion of beryllium (Be) plasma-facing components is explored, for the first time, using the ERO code. The code is applied to measured, RF-enhanced edge Be line emission at JET Be outboard limiters, in the presence of high-power, ion cyclotron-resonance heating (ICRH) in L-mode discharges. In this first modelling study, the RF sheath effect from an ICRH antenna on a magnetically connected, limiter region is simulated by adding a constant potential to the local sheath, in an attempt to match measured increases in local Be I and Be II emission of factors of 2–3. It was found that such increases are readily simulated with added potentials in the range of 100–200 V, which is compatible with expected values for potentials arising from rectification of sheath voltage oscillations from ICRH antennas in the scrape-off layer plasma. Absolute erosion values are also estimated within the uncertainties in local plasma conditions.

Stabilization of sawteeth with third harmonic deuterium ICRF-accelerated beam in JET plasmas

Sawtooth stabilisation by fast ions is investigated in deuterium (D) and D-helium 3 (He3) plasmas of JET heated by deuterium Neutral Beam Injection combined in synergy with Ion Cyclotron Resonance Heating (ICRH) applied on-axis at 3rd beam cyclotron harmonic. A very significant increase in the sawtooth period is observed, caused by the ICRH-acceleration of the beam ions born at 100 keV to the MeV energy range. Four representative sawteeth from four different discharges are compared with Porcelli's model. In two discharges, the sawtooth crash appears to be triggered by core-localized Toroidal Alfvén Eigenmodes inside the q = 1 surface (also called “tornado” modes) which expel the fast ions from within the q = 1 surface, over time scales comparable with the sawtooth period. Two other discharges did not exhibit fast ion-driven instabilities in the plasma core, and no degradation of fast ion confinement was found in both modelling and direct measurements of fast ion profile with the neutron camera. The developed sawtooth scenario without fast ion-driven instabilities in the plasma core is of high interest for the burning plasmas. Possible causes of the sawtooth crashes on JET are discussed.

Design of the 1.5 MW, 30-96 MHz ultra-wideband 3 dB high power hybrid coupler for Ion Cyclotron Resonance Frequency (ICRF) heating in fusion grade reactor.

Design and developmental procedure of strip-line based 1.5 MW, 30-96 MHz, ultra-wideband high power 3 dB hybrid coupler has been presented and its applicability in ion cyclotron resonance heating (ICRH) in tokamak is discussed. For the high power handling capability, spacing between conductors and ground need to very high. Hence other structural parameters like strip-width, strip thickness coupling gap, and junction also become large which can be gone upto optimum limit where various constrains like fabrication tolerance, discontinuities, and excitation of higher TE and TM modes become prominent and significantly deteriorates the desired parameters of the coupled lines system. In designed hybrid coupler, two 8.34 dB coupled lines are connected in tandem to get desired coupling of 3 dB and air is used as dielectric. The spacing between ground and conductors are taken as 0.164 m for 1.5 MW power handling capability. To have the desired spacing, each of 8.34 dB segments are designed with inner dimension of 3.6 × 1.0 × 40 cm where constraints have been significantly realized, compensated, and applied in designing of 1.5 MW hybrid coupler and presented in paper.

Real-time control of ELM and sawtooth frequencies: similarities and differences

ELMs and Sawteeth, located in different parts of the plasma, are similar from a control engineering point of view. Both manifest themselves through quiescent periods interrupted by periodic collapses. For both, large collapses, following long quiescent periods, have detrimental effects while short periods are associated with decreased confinement. Following the installation of the all metal ‘ITER like wall’ on JET, sawteeth and ELMs also play an important role by expelling tungsten from the core and edge of the plasma respectively. Control of tungsten has therefore been added to divertor heat load reduction, NTM avoidance and helium ash removal as reasons for requiring ELM and sawtooth control. It is therefore of interest to implement control systems to maintain the sawtooth and ELM frequencies in the desired ranges. On JET, ELM frequency control uses radial field ‘kicks’ and pellet and gas injection as actuators, while sawtooth control uses ion cyclotron resonance heating (ICRH). JET experiments have, for the first time, established feedback control of the ELM frequency, via real time variation of the injected gas flow []. Using this controller in conjunction with pellet injection allows the ELM frequency to be kept as required despite variations in pellet ELM triggering efficiency. JET Sawtooth control experiments have, for the first time, demonstrated that low field side ICRH, as foreseen for ITER, can shorten sawteeth lengthened by central fast ions []. The development of ELM and sawtooth control could be key to achieve stable high performance JET discharges with minimal tungsten content. Integrating such schemes into an overall control strategy will be required in future tokamaks and gaining experience on current tokamaks is essential.

Theoretical Analysis of Triple Liquid Stub Tuner Impedance Matching for ICRH on Tokamaks**supported by the National Magnetic Confinement Fusion Science Program of China (Nos. 2014GB108002, 2013GB107001), National Natural Science Foundation of China (Nos. 11205086, 11205053, 11375085, and 11405082), the Construct Program of Fusion and Plasma Physics Innovation Team in Hunan Province, China (No. NHXTD03), the Natural Science Foundation of Hunan

The impedance matching is crucial for continuous wave operation of ion cyclotron resonance heating (ICRH) antennae with high power injection into plasmas. A sudden increase in the reflected radio frequency power due to an impedance mismatch of the ICRH system is an issue which must be solved for present-day and future fusion reactors. This paper presents a method for theoretical analysis of ICRH system impedance matching for a triple liquid stub tuner under plasma operational conditions. The relationship of the antenna input impedance with the plasma parameters and operating frequency is first obtained using a global solution. Then, the relations of the plasma parameters and operating frequency with the matching liquid heights are indirectly obtained through numerical simulation according to transmission line theory and matching conditions. The method provides an alternative theoretical method, rather than measurements, to study triple liquid stub tuner impedance matching for ICRH, which may be beneficial for the design of ICRH systems on tokamaks.

On RF heating of inhomogeneous collisional plasma under ion-cyclotron resonance conditions

During ion-cyclotron resonance (ICR) heating of plasma by the magnetic beach method, as well as in some other versions of ICR heating, it is necessary to excite Alfven oscillations. In this case, it is difficult to avoid the phenomenon of the Alfven resonance, in which Alfven oscillations transform into lower hybrid oscillations. The latter efficiently interact with electrons, due to which most of the deposited RF energy is spent on electron (rather than ion) heating. The Alfven resonance takes place due to plasma inhomogeneity across the external magnetic field. Therefore, it could be expected that variations in the plasma density profile would substantially affect the efficiency of the interaction of RF fields with charged particles. However, the results obtained for different plasma density profiles proved to be nearly the same. In the present work, a plasma is considered the parameters of which correspond to those planned in future ICR plasma heating experiments on the PS-1 facility at the Kurchatov Institute. When analyzing the interaction of RF fields with charged particles, both the collisionless resonance interaction and the interaction caused by Coulomb collisions are taken into account, because, in those experiments, the Coulomb collision frequency will be comparable with the frequency of the heating field. Antennas used for ICR heating excite RF oscillations with a wide spectrum of wavenumbers along the magnetic field. After averaging over the spectrum, the absorbed RF energy calculated with allowance for collisions turns out to be close to that absorbed in collisionless plasma, the energy fraction absorbed by electrons being substantially larger than that absorbed by ions.

Cryogenic considerations for superconducting magnet design for the material plasma exposure experiment

In order to determine long term performance of plasma facing components such as diverters and first walls for fusion devices, next generation plasma generators are needed. A Material Plasma Exposure eXperiment (MPEX) has been proposed to address this need through the generation of plasmas in front of the target with electron temperatures of 1-15 eV and electron densities of 1020 to 1021 m-3. Heat fluxes on target diverters could reach 20 MW/m2. To generate this plasma, a unique radio frequency helicon source and heating of electrons and ions through Electron Bernstein Wave (EBW) and Ion Cyclotron Resonance Heating (ICRH) has been proposed. MPEX requires a series of magnets with non-uniform central fields up to 2 T over a 5-m length in the heating and transport region and 1 T uniform central field over a 1-m length on a diameter of 1.3 m. Given the field requirements, superconducting magnets are under consideration for MPEX. In order to determine the best construction method for the magnets, the cryogenic refrigeration has been analyzed with respect to cooldown and operational performance criteria for open-cycle and closed-cycle systems, capital and operating costs of these system, and maturity of supporting technology such as cryocoolers. These systems will be compared within the context of commercially available magnet constructions to determine the most economical method for MPEX operation. The current state of the MPEX magnet design including details on possible superconducting magnet configurations is presented.

Tokamak Plasma Flows Induced by Local RF Forces**supported by National Natural Science Foundation of China (Nos. 11405218, 11325524, 11375235 and 11261140327), in part by the National Magnetic Confinement Fusion Science Program of China (Nos. 2013GB111002, 2013GB112001 and 2013GB112010), and the Program of Fusion Reactor Physics and Digital Tokamak with the CAS “One-Three-Five” Strategic Planning

The tokamak plasma flows induced by the local radio frequency (RF) forces in the core region are analyzed. The effective components of local RF forces are composed of the momentum absorption term and the resonant parallel momentum transport term (i.e. the parallel component of the resonant ponderomotive forces). Different momentum balance relations are employed to calculate the plasma flows depending on different assumptions of momentum transport. With the RF fields solved from RF simulation codes, the toroidal and poloidal flows by these forces under the lower hybrid current drive and the mode conversion ion cyclotron resonance heating on EAST-like plasmas are evaluated.

The merits of ion cyclotron resonance heating schemes for sawtooth control in tokamak plasmas

JET experiments have compared the efficacy of low- and high-field side ion cyclotron resonance heating (ICRH) as an actuator to deliberately minimise the sawtooth period. It is found that low-field side ICRH with low minority concentration is optimal for sawtooth control for two main reasons. Firstly, low-field side heating means that any toroidal phasing of the ICRH ($-90^{\circ }$,$+90^{\circ }$or dipole) has a destabilising effect on the sawteeth, meaning that dipole phasing can be employed, since this is preferable due to less plasma wall interaction from Resonant Frequency (RF) sheaths. Secondly, the resonance position of the low-field side ICRH does not have to be very accurately placed to achieve sawtooth control, relaxing the requirement for real-time control of the RF frequency. These empirical observations have been confirmed by hybrid kinetic–magnetohydrodynamic modelling, and suggest that the ICRH antenna design for ITER is well positioned to provide a control actuator capable of having a significant effect on the sawtooth behaviour.

Use of virtual reality for optimizing the life cycle of a fusion component

Efficient development of a complex system such as a fusion component needs a stringent integration of standard and new constraints. For example, compared to the previous fusion experimental devices, remote handling (RH) and safety requirements are in ITER key parameters which must be integrated since the earliest design. For optimizing such integration studies, CEA, IRFM decided in 2010 to implement the use of virtual reality (VR) tools during the life cycle (from design to operation) of a fusion component. This paper describes a first feedback of such use for fusion engineering purposes.After a short overview of the CEA, IRFM VR platform capabilities, three main uses will be described: design review, simulation of remote handling and hands-on operations, with man in the loop.The Design review mode was intensively used within the framework of a fruitful collaboration with ITER design Integration Team. This mode, fully compatible with CAD software, enables scale one data visualization with stereoscopic rendering. It improves the efficiency in detecting inconsistencies inside models and machine sub-system design optimization needs. Several accessibility cases of major Safety Important Components (SIC-1) were studied giving important requirements to the design at an early stage.CEA, IRFM, in close collaboration with expertise of CEA, LIST for VR simulation software, applies VR technologies for designing RH maintenance scenario for ITER Test Blanket System (TBS) and Ion cyclotron Resonance Heating (ICRH) Port Plugs. RH compatibility studies using VR pointed out major design drivers while helping to propose credible solution. VR platform is intensively used in the design of WEST (Tungsten (W) Environment Steady-state Tokamak) components and assembly studies, providing important information about the feasibility of assembly processes, optimization of physical mock-ups and ergonomic posture and gestures of operator.Finally, new perspectives, as the integration of safety constraints (dose calculation) will be described, demonstrating the powerful of VR tools at different stages of the component lifecycle.

Physical program and diagnostics of the T-15 upgrade tokamak (brief overview)

Kurchatov Institute is upgrading now the T-15 tokamak to the machine with D-shaped plasma and copper magnetic system, capable for realizing lower and upper single-null and double-null magnetic configurations. The heating and current drive (CD) system consisting of the neutral beam injection (NBI), electron cyclotron resonance heating (ECRH/CD), electron Bernstein waves (EBW) heating and CD, ion cyclotron resonance heating (ICRH/CD), helicon and Lower Hybrid (LH) waves heating and CD is aiming to provide an effective heating of both electrons and ions, and on- and off-axis CD. The main research topics foreseen are the features of the confinement at high magnetic field and low aspect ratio, Advanced Tokamak regimes, steady-state operation, effects of turbulence with an emphasis on the role of the radial electric field Er, Geodesic Acoustic Modes (GAM) and Zonal Flows (ZF) in transport and confinement (including plasma self-organization, profile resiliency, influence of the q-profile), investigations of MHD effects and disruptions, Alfvén Eigenmodes (AE) and fast particles. Extended set of advanced diagnostics with identical equipment located at two toroidal positions will contribute to the 3D reconstruction of various types of the plasma structures like quasicoherent modes and long-range correlations.

Simulation as a tool to improve wave heating in fusion plasmas

Firstly, a brief overview will be given on different models that are able to describe the behaviour of wave propagation as a function of specific frequency ranges. Each range corresponds to different heating systems, namely, 20–100 MHz for the ion cyclotron resonant heating, 2–20 GHz for lower-hybrid heating or current drive, and 100–250 GHz for electron cyclotron resonant heating or current drive systems. The specification of every system will be explained in detail, including the typical set of equations and the assumptions needed to describe the properties of these heating or current drive systems, as well as their specific domains of validity. In these descriptions, special attention will be paid to the boundary conditions. A review of specific physical problems associated with the wave heating systems will also be provided. The review will detail the role of simulation in answering questions that arise from experiments on magnetized plasma devices devoted to fusion. A few examples that will be covered are the impact of edge turbulence on wave propagation and its consequences on heating system performance, the effects of fast particles and ponderomotive effects, among others. A study that is more focused on radio-frequency sheath effects will also be discussed. It shows that such simulations require very sophisticated tools to gain a partial understanding of the observations undertaken in dedicated experiments. To conclude this review, an overview will be given about the requirements and progress necessary to obtain relevant predictive simulation tools able to describe the wave heating systems used in fusion devices.

The mechanical structure of the WEST Ion Cyclotron Resonant Heating launchers

The WEST ICRH system has to deal with two challenging issues that no other ICRH system before ITER has faced simultaneously so far, i.e. ELMs resilience and Continuous Wave (CW) RF operation. The technical solution chosen to meet the requests imposed by the WEST scenarios is to build three new launchers based on the RF structure successfully tested in short pulses in 2007 on Tore Supra prototype launcher. This paper gives an overview of the mechanical structure of the CW ELMs resilient WEST ICRH launchers. The technical solutions chosen to drive the mechanical design are presented, in regard of the past experience on the 2007 TS prototype, together with the significant work carried out on the mechanical design to improve the launcher structure. The thermal and electro-mechanical analyses conducted and their impact on the launcher design are also presented. These three new CW ELMs resilient ICRH launchers are foreseen to be installed on WEST in 2016, and operational for the first plasmas.

Second-harmonic ion cyclotron resonance heating scenarios of Aditya tokamak plasma

Plasma heating with the fast magnetosonic waves in the ion cyclotron range of frequencies (ICRF) is one of the auxiliary heating schemes of Aditya tokamak. Numerical simulation of second-harmonic resonance heating scenarios in low-temperature, low-density Aditya plasma has been carried out for fast magnetosonic wave absorption in ICRF range, using full-wave ion cyclotron heating code TORIC combined with Fokker–Planck quasilinear solver SSFPQL and the results are explained. In such low-temperature, low-density plasma, ion absorption for second-harmonic resonance heating is less but significant amount of direct electron heating is observed.

Tungsten impurity transport experiments in Alcator C-Mod to address high priority research and development for ITERa)

Experiments in Alcator C-Mod tokamak plasmas in the Enhanced D-alpha H-mode regime with ITER-like mid-radius plasma density peaking and Ion Cyclotron Resonant heating, in which tungsten is introduced by the laser blow-off technique, have demonstrated that accumulation of tungsten in the central region of the plasma does not take place in these conditions. The measurements obtained are consistent with anomalous transport dominating tungsten transport except in the central region of the plasma where tungsten transport is neoclassical, as previously observed in other devices with dominant neutral beam injection heating, such as JET and ASDEX Upgrade. In contrast to such results, however, the measured scale lengths for plasma temperature and density in the central region of these Alcator C-Mod plasmas, with density profiles relatively flat in the core region due to the lack of core fuelling, are favourable to prevent inter and intra sawtooth tungsten accumulation in this region under dominance of neoclassical...

R&D activities on RF contacts for the ITER ion cyclotron resonance heating launcher

Embedded RF contacts are integrated within the ITER ICRH launcher to allow assembling, sliding and to lower the thermo-mechanical stress. They have to withstand a peak RF current up to 2.5kA at 55MHz in steady-state conditions, in the vacuum environment of the machine. The contacts have to sustain a temperature up to 250°C during several days in baking operations and have to be reliable during the whole life of the launcher without degradation. The RF contacts are critical components for the launcher performance and intensive R&D is therefore required, since no RF contacts have so far been qualified at these specifications. In order to test and validate the anticipated RF contacts in operational conditions, CEA has prepared a test platform consisting of a steady-state vacuum pumped RF resonator. In collaboration with ITER Organization and the CYCLE consortium (CYclotron CLuster for Europe), an R&D program has been conducted to develop RF contacts that meet the ITER ICRH launcher specifications. A design proposed by CYCLE consortium, using brazed lamellas supported by a spring to improve thermal exchange efficiency while guaranteeing high contact force, was tested successfully in the T-resonator up to 1.7kA during 1200s, but failed for larger current values due to a degradation of the contacts. Details concerning the manufacturing of the brazed contacts on its titanium holder, the RF tests results performed on the resonator and the non-destructive tests analysis of the contacts are given in this paper.

Design and optimization of the WEST ICRH antenna front face components based on thermal and hydraulic analysis

The WEST (Tungsten (W) Environment in Steady-state Tokamak) is an upgrade of Tore-Supra (TS) which aims it into an X-point magnetic configuration tokamak equipped with an actively cooled tungsten divertor. To be a platform of ITER technologies of high heat flux components testing, three sets of Ion Cyclotron Resonant Heating (ICRH) antennas have been designed to inject 9MW during 30s or 3MW during 1000s.The antenna design is based on a load resilient prototype successfully tested in Tore Supra in 2007. In order to allow continuous-wave (CW) operations, the mechanical design of the WEST ICRH antenna is emphasized on its cooling performances by designing fully active cooling structure. Two kinds of cooling water loops are used, with temperature and pressure of 70°C/30bar and 25°C/5.2bar, respectively. The hot water loop is used for the Faraday screen (FS) and the housing box (HB), while the cold water loop is used for the straps, the matching capacitors and the impedance transformer. To enhance the heat removal ability and control the pressure drop, the cooling channels in the FS and HB are drilled directly and parallel connected as much as possible.By performing the hydraulic–thermal analysis, the lack of cooling efficiency was found in the front face of lateral collector where 1MW/m2 is imposed and fluid dead zones were found in some of the bars. After optimization, the cooling performance of the cooling circuit increased significantly. With a mass flow rate of 2.5kg/s, the total pressure drop is 3.1bar, and the peak temperatures on the FS and HB are 500°C and 261°C, respectively. Besides, no cavitation is found in the total cooling circuit.