Electron Cyclotron Heating System Research Articles

Measurements of microwave transmission mode in HL-3 tokamak relevant ECH corrugated waveguides transmission line

The transmission modes of high-power millimeter wave in corrugated waveguides transmission line (TL) are crucial, especially for the long-pulse operation for long-distance TL. The microwave transmission mode in the TL on HL-3 tokamak relevant the electron cyclotron heating system are studied for the purpose according to the phase retrieval method based on intensity distribution measured by an infrared camera at several different locations away from the outlet of TL. A specific high power test platform with about 40-meter TL is established and the mode contents in the middle and the end of the TL are analyzed. In the middle of the TL, the proportion of main transmission mode LP01 mode is about 97% and other high-order modes contents are less than 3%. In the end of the TL, the main mode purity is about 94%, and the high-order mode, LP11 even mode is excited about 4.6%. These mode analysis results indicate that the high-power and long-pulse millimeter wave corrugated waveguides TL are designed and machined well, and the alignment method for long distance transmission line installation is highly useful for high-purity wave mode transmission.

Conceptual design of a modular EC heating system for EU-DEMO

The European DEMO (EU-DEMO) reactor studies within EUROfusion aim to develop a fusion power plant concept. The large tokamak device needs an auxiliary heating power which, at the present stage, is provided by the Electron Cyclotron (EC) heating system with up to 130 MW foreseen to reach different regions of plasma for heating, suppression of instabilities and the possibility to support ramp-up and ramp-down phases. The present conceptual design of the system is based on 2 MW coaxial-cavity gyrotron sources, a transmission line (TL) using both circular corrugated waveguides and quasi-optical evacuated multi-beam TLs, and mirror antennas located in the Equatorial Port. In order to create a modular system, the sources are grouped in ‘clusters’, whose powers are combined in the quasi-optical TL, up to the tokamak building, where they are split and routed as single waveguides. In the launcher, they are combined together again on the launching mirrors, to save space for the apertures in the Breeding Blanket. The present EC heating system has a certain flexibility to adapt to changing design guidelines. The development status of the system is presented.

Divertor-safe nonlinear burn control based on a SOLPS parameterized core-edge model for ITER

For ITER operations, the range of desirable burning-plasma regimes with high fusion power output will be restricted by various operational constraints. These constraints include the saturation of ITER’s various heating and fueling actuators such as the neutral beam injectors, the ion and electron cyclotron heating systems, the gas puffing system, and the deuterium–tritium pellet injectors. In addition to these actuator constraints, the H-mode power threshold, divertor detachment, and the heat load on the divertor targets may apply limitations to ITER’s operational space. In this work, Plasma Operation Contour (POPCON) plots that map the aforementioned constraints to the temperature-density space are used to investigate which constraints are most limiting towards accessing regimes with high fusion power output. The presented POPCON plots are based on a control-oriented core-edge model that couples the nonlinear density and energy response models for the core-plasma region with SOLPS4.3 parameterizations for conditions in the edge-plasma regions (scrape-off-layer and divertor). Using this control-oriented core-edge model, a nonlinear burn controller, which aims to regulate the plasma temperature and density in the core-plasma region, is constructed in this work. This controller is augmented with an online optimization scheme that governs the control references such that the plasma can be guided towards regimes with high fusion powers while protecting the divertor targets from dangerously high heat loads. A closed-loop simulation study illustrates the capability of this burn control scheme.

Design and operation results of KSTAR ECH system

The electron cyclotron heating (ECH) system at the Korea Superconducting Tokamak Advanced Research (KSTAR) is a crucial heating devices, which is employed in various research endeavors, including plasma startup, wall cleaning, neoclassical tearing mode control, long pulse operation, and hybrid scenarios development from the first plasma operation. For facilitating advanced scenarios development and magnetohydrodynamics research in KSTAR, it was deemed necessary to install a total of 6 MW ECH systems, consisting of four 105/140 GHz and two 170 GHz systems. To address this, a new ECH room was developed and the existing ECH system was totally rearranged. This study describes the design of the novel KSTAR ECH system and its operation results, and discusses the problems that occur during the ECH system operation. Each ECH system to deliver 1 MW for 300 s was designed to enable stable, safe, and efficient operation in the new ECH room and four 105/140 GHz dual frequency ECH were injected to KSTAR successfully. The polarization angle was experimentally confirmed by using an electron cyclotron wall conditioning, and the beam angle change by ECH antenna was verified by a newly developed IR measurement. However, mixed patterns with higher-order modes were observed in the output beam from matching optics unit of some gyrotrons, which affect the operation efficiency and the output power. In future works, we will analyze the causes for low mode purity and improve the operation efficiency.

Progress and challenges of the ECH transmission line design for DTT

The design of the Transmission Line (TL) as a part of the Electron Cyclotron Heating (ECH) system for Divertor Tokamak Test facility (DTT) is approaching the conceptual design maturity. With an ECH system of 16 MW installed for the first phase and with a total of 32 gyrotrons (170 GHz, ≥ 1 MW, 100 s) the TL design is undertaking the challenge of an evacuated Multi-Beam TL (MBTL) concept to deliver the large number of beam lines from the gyrotron hall to the torus hall buildings. The system is organized in 4 clusters, each of them including 8 beamlines. The routing consists of single-beam TL section used to connect the gyrotron output to a beam-combiner mirror unit for each cluster, a common MBTL running in a suspended corridor reaching the Tokamak building and a beam-splitter mirror unit to connect to the ex-vessel optics and launchers sections located in the equatorial and upper ports of one sector, for a total of 4 DTT sectors. The TL mirrors will be actively water cooled to cope with the heat load in long pulses due to the high power incident radiation, with the possibility to include advanced concepts for the cooling design compatible with additive manufacturing technology. The characteristics of the system and its components are presented, showing both the progress of the adopted solutions and the current design. Since the main challenge of this TL is to maintain the overall losses below 15%, in this paper we present the expected ohmic and spillover losses, including beam coupling simulations evaluating losses given by high order Transverse Electro-Magnetic modes (i.e. aberrations). We describe how the effects have been estimated with electromagnetic simulations and how losses could be mitigated, since TL efficiency could significantly drop due to the presence of non-idealities, like the deformations of mirrors surface ascribed to the microwaves heat loads and possible misalignments and aberrations effects occurring along the line.

Pre-conceptual design of the steering mirror for the DEMO electron cyclotron heating system

An Electron Cyclotron (EC) Heating and Current Drive (HCD) system will be installed in the EU DEMO Tokamak to provide plasma heating and stabilization by coupling up to 130 MW of mm-wave power into specific regions of the plasma. The mm-waves will be generated by gyrotrons and propagated through a quasi-optical multi-beam mirror system (MBMS), and subsequently through corrugated waveguides, to launchers located at dedicated Equatorial Ports (EPs). The present design of the launcher features two mid steering antennas (MSAs) for plasma stabilization each consisting of three-beam layouts and steering mirrors positioned at the upper and lower locations of the EP. The proposed concept for the steering mirrors operates on the pantograph principle, a mechanical assembly of parallel linkages, which translates the movement produced by pneumatic actuators into an exact angular rotation of the mirror around an offset virtual axis. This design uses flexures with specific structural compliance as mechanical joints resulting in the desired movement of the linkage system while avoiding the tribological difficulties associated with mechanisms operating under vacuum. This paper describes the present status of the DEMO EC HCD steering mirror as well as the analyses performed to validate this design.

Material Plasma Exposure eXperiment (MPEX) High Heat Flux Bellows for Component Alignment in High Microwave Environment

The Material Plasma Exposure eXperiment (MPEX) device is a linear plasma device developed to perform plasma material interaction experiments under the conditions prototypic of a fusion reactor divertor. MPEX has multiple systems that must be precisely aligned to the plasma axis, including an electron cyclotron heating system that emits up to 400 kW of microwave power into the vacuum vessel. Five distinct systems require precise alignment on the MPEX device, thus requiring four bellows, all of which are adjacent to the plasma at a relatively high heat flux of approximately 47 kW/m2 and microwave power regions. The MPEX high heat flux bellows (HHFB) is designed to deliver 6 degrees of freedom positioning. The HHFB includes titanium–zirconium–molybdenum (TZM) inserts that are brazed into a Glidcop AL-15 body using a high-temperature braze alloy, thus blocking direct line of sight to an edge-welded bellows from the plasma and microwave screen to block microwaves from the bellows. A custom ConFlat knife edge is machined into the Glidcop AL-15, so the vacuum flanges do not need a braze or weld joint on the vacuum interface. Fingerstock or copper mesh is used to restrict microwave power from entering the interstitial space between the water-cooled Glidcop AL-15 body and edge-welded bellows. Glidcop AL-15 was selected as the material of choice for the water-cooled body because it can maintain mechanical integrity at elevated temperatures, and it also allows for a high-temperature braze. TZM was selected for its machinability and compatibility with vacuum and plasma requirements. A test article is also being considered for a similarly shaped component, the MPEX limiter, that will demonstrate the integrity of the braze joint under high thermal load. Results from this testing will be extrapolated to deduce the lifetime and integrity of the HHFB design.

Nonlinear burn control in ITER using adaptive allocation of actuators with uncertain dynamics

ITER will be the first tokamak to sustain a fusion-producing, or burning, plasma. If the plasma temperature were to inadvertently rise in this burning regime, the positive correlation between temperature and the fusion reaction rate would establish a destabilizing positive feedback loop. Careful regulation of the plasma’s temperature and density, or burn control, is required to prevent these potentially reactor-damaging thermal excursions, neutralize disturbances and improve performance. In this work, a Lyapunov-based burn controller is designed using a full zero-dimensional nonlinear model. An adaptive estimator manages destabilizing uncertainties in the plasma confinement properties and the particle recycling conditions (caused by plasma–wall interactions). The controller regulates the plasma density with requests for deuterium and tritium particle injections. In ITER-like plasmas, the fusion-born alpha particles will primarily heat the plasma electrons, resulting in different electron and ion temperatures in the core. By considering separate response models for the electron and ion energies, the proposed controller can independently regulate the electron and ion temperatures by requesting that different amounts of auxiliary power be delivered to the electrons and ions. These two commands for a specific control effort (electron and ion heating) are sent to an actuator allocation module that optimally maps them to the heating actuators available to ITER: an electron cyclotron heating system (20 MW), an ion cyclotron heating system (20 MW), and two neutral beam injectors (16.5 MW each). Two different actuator allocators are presented in this work. The first actuator allocator finds the optimal mapping by solving a convex quadratic program that includes actuator saturation and rate limits. It is nonadaptive and assumes that the mapping between the commanded control efforts and the allocated actuators (i.e. the effector model) contains no uncertainties. The second actuator allocation module has an adaptive estimator to handle uncertainties in the effector model. This uncertainty includes actuator efficiencies, the fractions of neutral beam heating that are deposited into the plasma electrons and ions, and the tritium concentration of the fueling pellets. Furthermore, the adaptive allocator considers actuator dynamics (actuation lag) that contain uncertainty. This adaptive allocation algorithm is more computationally efficient than the aforementioned nonadaptive allocator because it is computed using dynamic update laws so that finding the solution to a static optimization problem is not required at every time step. A simulation study assesses the performance of the proposed adaptive burn controller augmented with each of the actuator allocation modules.

Preparation for assembly and commissioning of ITER

Progress is described on the construction of infrastructure and plant systems, and on the manufacturing of the ITER tokamak components and their assembly since the IAEA FEC 2018. Major achievements in this period have been the completion of the tokamak building and installation of the cryostat base and lower section that mark the start of tokamak assembly in 2020. This has been followed by the arrival of key core tokamak components to enable assembly of the first tokamak sector, which is made up of one vacuum vessel (VV) sector, 2 toroidal field (TF) coils and the corresponding thermal shield (TS), and the completion of manufacturing and testing of the first two poloidal field (PF) coils (PF5 and PF6). In parallel with this, the tools required for assembly have been commissioned and plant systems required for operations have progressed in construction and have started or will start commissioning in the near future. The remaining core tokamak components (e.g. the remaining eight VV sectors, TF coils, PF coils and central solenoid (CS) modules) are in advanced manufacturing state or already completed and under testing, as required for the assembly plan to first plasma (FP). In parallel with machine assembly, systems to support plasma operation during the FP and later phases of the ITER research plan (IRP) have made significant progress as well. This includes the plasma control system, for which the final design for FP has been completed, the electron cyclotron heating (ECH) system, diagnostic systems, plasma protection components for FP and the disruption mitigation system (DMS). The IRP itself has also been consolidated and specific issues identified where R&D support by the ITER Members’ fusion communities may refine it further. The consolidation of the IRP has involved changes to the upgrade paths considered in the baseline by the identification of MHD stable plasma scenarios that can meet ITER’s Q = 5 steady-state fusion power goal with neutral beam heating (NBI) and ECH thus removing the need for the installation of a lower hybrid current drive (LHCD) system previously included as an upgrade in the baseline for this specific purpose.

Development of a 56 GHz ECH system for deuterium plasma experiments of a low magnetic field in LHD

We have completed establishing a 56GHz electron cyclotron heating (ECH) system with a new gyrotron to realize high β plasma experiments with pure deuterium gas in the Large Helical Device (LHD). This new ECH system made it possible to conduct the relatively low magnetic field experiment around 1T with pure deuterium gas because tangential neutral beam injection (NBI) systems of deuterium beams did not have enough power to initiate plasma in the LHD. We succeeded in initiating and sustaining pure deuterium plasma in the magnetic field of 1T by using the 56GHz ECH system and the deuterium NBI (D-NBI). This new ECH system contributed to expanding the new experimental regime in the LHD.

Structural Analysis of the US ITER ECH Transmission Line System

The electron cyclotron heating (ECH) and current drive system is one of the main plasma heating systems for ITER. It uses high-power microwave beams with the power deposition location steerable across the plasma cross section. Microwave power is conveyed via transmission lines (TLs) that run from the gyrotrons in the radio frequency building through the assembly hall and tokamak building to the ECH launchers within the tokamak vacuum vessel. The ECH system includes a vast array of interconnected TL waveguides, in-line components, and support structures. Finite element (FE) modeling provides an essential means of simulating the system, applying loads and determining deflections, rotations, forces, moments, and stresses in order to evaluate various structural and microwave transmission performance metrics. A representative FE model of the overall ECH TL system is developed in ANSYS®. This top-level model defines the centerline of the waveguide system. Waveguide segments are represented by line elements (beams and pipes) with equivalent section properties, and support structures are represented by boundary conditions. A systematic approach is used to model each ECH component with lumped masses and structurally equivalent stiffness matrices or ANSYS superelements. The top-level TL FE model is used to evaluate the various loads (thermal, vacuum, seismic, etc.) and operating scenarios. The top-level model directly calculates stresses in the straight aluminum waveguide segments. The model provides the forces and moments acting on the in-line components for detailed submodel assessments. Displacement results from the top-level analysis feed into a separate microwave performance model to help determine operational efficiency. All TL performance and thermal-structural requirements are met, as specified by the applicable codes and standards, and successfully documented in numerous technical reports and demonstrated at the final design review.

CVD diamond photodetectors for FTU plasma diagnostics

We report on the results achieved by installing two Chemical Vapor Deposition (CVD) single crystal diamond detectors in one of the equatorial ports of the Frascati Tokamak Upgrade (FTU) tokamak, during the last six months of operation of the machine. The devices were fabricated at the University of Rome “Tor Vergata” in a metal/instrinsic/p-type diamond layered structure, allowing them to be used as Schottky photodiodes for VUV and Soft-X ray detection. Both detectors were placed inside the machine and operated in current mode under high vacuum conditions.The fast response capabilities of diamond detectors allowed to observe several plasma events, like the so-called Anomalous Doppler instabilities, the pellet injection and ablation, and Multifaceted Asymmetric Radiation From the Edge (MARFE). Diamond detectors often but not always followed the Magnetohydrodynamics (MHD) activity, depending on their localization relative to the emitting region. Core temperature oscillations following Electron Cyclotron Heating system (ECH) modulation were also observed. In addition, the diamond signals were compared to selected channels of the FTU bolometry system with similar line-of-sight: the encouraging results have launched an R&D program for the development of diamond-based bolometers.

Fluid-dynamic and thermo-mechanical analyses of the Monoblock Mitre Bend for the ITER electron cyclotron Upper launcher

The ITER Electron Cyclotron Heating (ECH) system will be used to counteract magneto-hydrodynamic plasma instabilities by aiming up to 20 MW of mm-wave power at 170 GHz. The primary vacuum boundary at the Electron Cyclotron Upper Launcher (EC UL) extends into the port cell region through eight beamlines, defining the so-called First Confinement System (FCS). Each beamline, designed for the transmission of 1.31 MW, is delimited by the closure plate at the port plug back-end and by a diamond window in the port cell. The FCS essentially consists of a Z-shaped set of straight corrugated waveguides (WG) connected by Mitre Bends (MB) with a nominal inner diameter of 50 mm. Due to the space restrictions, the eight MBs at the last FCS section are grouped into two monoblocks. Each Monoblock Mitre Bend (MBMB) consists of a body with corrugated feedthroughs defining a specific angle for each beamline and four mirrors attached by bolted connection to reflect the mm-wave power. The thermal expansion arising from the ohmic losses, the cooling pressure, the bolt pre-tension and the imposed displacements coming from the connection with the transmission lines are the primary design loads for the MBMBs. The fluid-dynamic analyses performed for both upper and lower MBMBs show that highest temperature takes place in the MB mirrors, reaching a maximum value of 203 °C at the beam center. The thermo-mechanical analyses demonstrate that the peak stress also occurs at this location with a maximum value of 324 MPa. These stresses are categorized and compared with the allowable limits defined in the ASME code, what probes that the current design of both upper and lower MBMBs are able to withstand the loads taking place during the mm-wave normal operation.

The Materials Plasma Exposure eXperiment: Status of the Physics Basis Together With the Conceptual Design and Plans Forward

The Materials Plasma Exposure eXperiment (MPEX) is a linear plasma device that will address the plasma–material interaction (PMI) science for future fusion reactors and will enable testing of plasma-facing components (PFCs). It is designed as a steady-state device eventually delivering an ion fluence of up to 1031 m−2 to the target. The device will be designed to handle neutron-activated materials. These capabilities should push the technical readiness level of PFCs up to six for some end-of-life aspects. In order to achieve the relevant plasma conditions, as they are expected in future fusion reactor divertors, MPEX will utilize a novel plasma source concept. This plasma source concept uses a high-power helicon (200 kW, 13.56 MHz), an electron cyclotron heating (ECH) system, which will heat electrons via electron Bernstein wave (EBW) heating (about 400 kW, 70 GHz), and an ion cyclotron heating (ICH) system (400 kW, 6–9 MHz). The physics basis for this plasma source concept including the heating physics and transport is based on experiments on Proto-MPEX. An overview of the experimental results and the physics basis will be given. The physics basis directly translates into functional requirements of MPEX and the conceptual design. The status of the conceptual design of MPEX will be shown.

Design of Central Controller for EAST ECRH System Based on FPGA

Electron cyclotron resonance heating is one of the important heating methods for controlled nuclear fusion reactions, which can make up for the shortcomings of the traditional heating methods. To this end, a 4MW long pulse electron cyclotron heating system was designed on the Experimental Advanced Superconducting Tokamak device. This paper mainly introduces the design of the gyrotron controller of ECRH system based on FPGA chip. The control logic of the central controller is based on state machine. Throughout the entire control process, each signal in the system is monitored in real time. Once an abnormality occurs, the main control logic will cut off the high voltage power of the gyrotron as required to protect the gyrotron from damage. In addition to the normal control logic, the FPGA also embeds abnormality diagnosis logic, which monitors the control state transitions of the main control logic and various input signals in real time to determine whether an abnormality has occurred in the system and can diagnose the cause of the abnormality. Timing simulation was done to the FPGA program. The result shows that the controller can respond to the total control trigger signal in real time, accurately control the switching of the cathode and anode high voltage power supply, diagnose and handle various abnormal conditions, and meet the control requirements of the ECRH system.

Control of the gyrotron filament power supplies for the experimental advanced superconducting tokamak.

The electron cyclotron heating system including four gyrotrons is being developed in the Institute of Plasma Physics, Chinese Academy of Sciences. The filament is an important part of the gyrotron, which is used to heat the cathode. The gyrotron output power can be controlled by adjusting the filament power. We have developed the filament power control and measurement system using Labview. The filament characteristics were tested using this power control system. The test results show that the filament current can be fitted with the e-exponential function of the filament voltage. It can be seen that as the voltage increases, the filament resistance gradually increases. We have developed a new "burst" function to prevent the beam current from dropping too fast, which is conducive to long pulse stable operation of the gyrotrons.

Equivalent Model of Transient Gyrotron Cathode Response

The gyrotron is the high power millimeter wave source in the electron cyclotron resonance heating system for the tokamak. The cathode power supply is one of the most important ancillary devices for gyrotron. Some interesting transient phenomena about the cathode voltage and the cathode current was found in the gyrotron operation in the electron cyclotron resonance heating system for the Experimental Advanced Superconducting Tokamak. The cathode voltage drops to about 10% of the original value by about 90 {\mu}s in the overcurrent case, which is much longer than the 25 {\mu}s in the normal case. In order to explain these phenomena, an equivalent circuit model of the magnetron injection gun was proposed. The equivalent circuit is composed of parallel resistors and capacitors, and it can explain the test results very well. Using the equivalent circuit model to analyze gyrotrons may become an effective means of developing and operating an electron cyclotron heating system.

QUEST における28 GHz 電子サイクロトロン加熱システムを用いた非誘導プラズマ電流立ち上げ実験

QUEST における28 GHz 電子サイクロトロン加熱システムを用いた非誘導プラズマ電流立ち上げ実験

Electrical integration of two 1MW/2s dual-frequency gyrotrons into the EC-system of the TCV tokamak

The electron-cyclotron heating system of the TCV tokamak is being enhanced as part of a major upgrade. Two MW-class dual-frequency gyrotrons will be installed. The first has already been delivered and will be fully commissioned on-site by the end of 2018 while the second is due for delivery in 2019.This paper describes the electrical integration of these two gyrotrons. The water-cooling system is also described from the electrical point of view. Results of initial measurements on the first gyrotron are described.Based on long term experience at the Swiss Plasma Center with gyrotrons, giving special emphasis to electromagnetic compatibility, overall electrical design choices will ensure reliable operation of the gyrotrons.

New features of the ECH control system on DIII-D and impact of power modulation on collector loading

In order to increase the reliability and improve the availability of electron cyclotron heating (ECH) and current drive during a plasma shot, an adaptive gyrotron command generator has been designed and implemented in all gyrotron systems on the DIII-D tokamak. This adaptive real-time generator is able to detect imminent faults, including gyrotron body overcurrent faults that can develop in a few microseconds, block faults, and restart RF generation within 10–30 ms, faster than the DIII-D energy confinement time. The adoption of this adaptive generator has substantially improved the performance of a depressed-collector gyrotron with recurrent body current faults, and has assisted the ECH system to achieve a record reliability of 95% during the last experimental campaign. The impact of power modulation on the collector loading of a depressed-collector gyrotron has been experimentally studied using calorimetry and collector maps taken by a diagnostic developed in-house. While lowering the modulation duty cycle kept the collector power constant as a result of low beam-wave efficiency during the modulation low state, an observed power redistribution inside the collector increased the peak collector power density, which reached levels above the manufacturer power density limit for low duty cycles.