Fusion Experimental Device Research Articles

Hydrogen interactions with intrinsic point defects and hydrogen diffusion in tungsten doped Al2O3

In light of the extensive use of tungsten (W) in fusion experimental devices and the importance of hydrogen isotopes permeation, based on the first principle approaches we study the effects of W doping on the formation energies of intrinsic point defects, hydrogen interactions with such defects, and hydrogen diffusion in bulk of α-Al2O3. Our investigation demonstrated that W doping enhances the formation of oxygen and aluminum vacancies (Vo and VAl) in α-Al2O3. In W doping α-Al2O3, hydrogen primarily exists in the form of Ho− (hydrogenation of the bulk VO3−) and Hi− (H interstitial) defects, while it hardly occupies VAl. Hydrogen is firmly trapped by Vo with a formation energy −3.35eV, and Hi− diffusion become extremely challenging because W doping significantly raises the Hi− diffusion barrier to 4.44 eV. Our findings indicate that W doping is beneficial for hydrogen permeation resistance in α-Al2O3.

Novel identification algorithm for plasma boundary gap based on visible endoscope diagnostic on EAST tokamak

The precise plasma boundary gap identification at the midplane is a prerequisite for achieving controlled plasma positioning and holds a significant importance for the stable operation of tokamak devices. This study proposes a plasma boundary gap at the midplane recognition algorithm based on visual endoscopy diagnostic. The model is an end-to-end one that uses a convolutional neural network that does not require manual data labeling. The model performance is improved by experimentally comparing different convolutional layers and input image sizes. The model is validated using a testing dataset comprising 400 plasma discharge moments. The model has average errors of 3.7 and 4 mm for gap-in and -out, respectively, when compared to those obtained by equilibrium fitting. The proposed approach offers a convenient and effective means of obtaining the boundary gap value and is particularly suited for future fusion experimental devices, such as BEST and ITER tokamak.

Compact fusion blanket using plasma facing liquid Li-LiH walls and Pb pebbles

Liquid plasma facing walls allow for increased neutron-wall loading expanding the design space of fusion power-plants and experimental devices towards compact high-field reactors. This study presents the design of a compact radial build blanket for fusion devices composed of variable quantities of Lead (Pb) and Lithium-Lithium Hydride (Li-LiH). A tank-like cylindrical neutronic model of the early design of the stellarator reactor proposed by Renaissance Fusion is implemented in OpenMC (neutron transport and dose rate analyses). The reactor's radial build composition and blanket layer thicknesses are varied to fulfill the requirements on tritium breeding ratio (TBR), nuclear heat extraction, radiation shielding (for the coils, internal structures and external environment) for a stellarator-based power-plant. The analyses suggest that a radial build lower than a meter thick between the plasma and coils would be sufficient to allow for a TBR ∼ 1.60, an energy multiplication factor of ∼ 1.07, to capture ≥ 90% of the nuclear heat, limit the neutron fluence at the coils below 1019 n/cm2, and limit the structural damage on the liquid metal vessel and magnet structure. In particular, a blanket composed of 32 cm of Pb and Li-LiH, 54 cm of a heavy metal hydride such as vanadium hydride (VH2), along with a 1.3 m of concrete bioshield, would minimize the radial build of the stellarator reactor while fulfilling tritium breeding, shielding and heat extraction requirements.

Design and operation of a load-tolerant ICRH system in Experimental Advanced Superconducting Tokamak

Ion Cyclotron Resonance Heating (ICRH) has been a dependable tool for sturdy plasma heating with high RF power of several megawatts. However, a sudden increase in the reflected power during ICRH heating experiments is a problem that should be solved for future fusion experimental devices. To solve this issue, a load tolerant matching network has been designed for the ICRH system in EAST. The matching network includes a 3-stub tuner impedance matching system with conjugate-T structure, 30 Ω to 50 Ω transmission line and center grounded antenna strap. By maintaining a low reflection ratio in the network for a wide range of resistance, this matching network can allow sturdy high-power operations without fast impedance matching in EAST. In our matching network, the two arms of a conjugate-T are designed to have a λ/2 length difference which could mitigate current imbalance and antenna poloidal phasing outside of the control problem. The T-point corresponds to the maximum point of the standing wave voltage, which could greatly improve the input impedance of the antenna.

Deep Learning Models to Reduce Stray Light in TJ-II Thomson Scattering Diagnostic.

Nuclear fusion is a potential source of energy that could supply the growing needs of the world population for millions of years. Several experimental thermonuclear fusion devices try to understand and control the nuclear fusion process. A very interesting diagnostic called Thomson scattering (TS) is performed in the Spanish fusion device TJ-II. This diagnostic takes images to measure the temperature and density profiles of the plasma, which is heated to very high temperatures to produce fusion plasma. Each image captures spectra of laser light scattered by the plasma under different conditions. Unfortunately, some images are corrupted by noise called stray light that affects the measurement of the profiles. In this work, we propose the use of deep learning models to reduce the stray light that appears in the diagnostic. The proposed approach utilizes a Pix2Pix neural network, which is an image-to-image translation based on a generative adversarial network (GAN). This network learns to translateimages affected by stray light to images without stray light. This allows for the effective removal of the noise that affects the measurements of the TS diagnostic, avoiding the need for manual image processing adjustments. The proposed method shows a better performance, reducing the noise up to 98% inimages, which surpassesprevious works that obtained 85% for the validation dataset.

Data-driven fault diagnosis method for abnormal RF oscillation of gyrotrons

Gyrotrons are vacuum electronic devices capable of generating high-power, high-frequency THz radiation. With the increasing utilization of gyrotrons in thermonuclear fusion experimental devices, achieving higher output performance and ensuring stable and reliable operation have become key development goals. However, occasional radio frequency (RF) oscillation faults inside the gyrotron during long-pulse operation hinder its stable and reliable operation. Due to the difficulty in directly observing abnormal changes within the gyrotron resonator during operation, this paper explores a data-driven approach to diagnose RF oscillation faults in the gyrotron for the first time and proposes a classification model that combines K-Nearest Neighbors and Random Forest (KNN-RForest) algorithms for fault identification. Compared with seven baseline models, the results show that our proposed KNN-RForest model has a better classification performance. It is verified that data-driven methods can effectively identify RF oscillation faults in gyrotrons. Finally, the state probabilities of the gyrotron under different power levels were predicted using the KNN-RForest model, demonstrating the application of the KNN-RForest model in identifying gyrotron RF oscillation faults, which may be helpful for setting stable operating parameters for the gyrotron to reduce the likelihood of RF oscillation faults.

Development of a Compact HTS-FAIR Conductor for Magnet Application

Research and development of large-current capacity high temperature superconducting (HTS) conductors for magnet applications in nuclear fusion experimental devices is underway in the world. As a candidate conductor, the FAIR conductor has been developed, in which REBCO tapes and pure aluminum sheets are alternately stacked in an aluminum jacket, and the lid is covered by friction stir welding (FSW) method. Producing several prototypes of FAIR conductors, current transport tests have been repeatedly conducted in liquid nitrogen without an external magnetic field to optimize the manufacturing process of the conductor. In addition, we have also investigated the properties of the conductors in low temperature under high magnetic fields. As the next step, we have been planning to conduct coil shape tests in high magnetic fields. For installation in a test facility, we have developed a reduced diameter FAIR conductor from 12 mm to 6 mm. 1.5 rotations/m reduced version of FAIR was fabricated and current transport tested in 77 K, self-magnetic field. Compared with the results of critical current analysis considering the self-magnetic field, the retention of <italic xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">I</i> <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">c</sub> was 67%. After cutting the fabricated the compact FAIR conductor and observing the inside, it was found that the REBCO tapes were folded by FSW. Therefore, it was found that further optimization of the FSW process was necessary.

Development of Directional 14 MeV-Fusion Neutron Detector Using Liquid-Scintillator-Filled Capillaries

Triton burnup studies have been performed in fusion experimental devices to evaluate the confinement performance of 3.5 MeV alpha particles. For this purpose, the generation rate of deuterium-tritium (D-T)-born 14 MeV neutrons has been measured using a scintillating fiber detector. The directionality of this detector provides excellent selectivity for 14 MeV neutrons; however, the lack of a pulse–shape discrimination (PSD) capability limits high-energy gamma-ray reduction. In this study, we developed a new 14 MeV neutron detector with directionality that can discriminate neutrons and gamma rays based on the PSD technique by filling capillaries with a liquid scintillator. The performance of this detector was evaluated at the FNL (Tohoku University, Japan) and OKTAVIAN (Osaka University, Japan). The detector response was modeled using the particle and heavy ion transport code system (PHITS). The experimental and simulation results demonstrated that the detector has a directional response to fast neutrons and excellent PSD capability.

Stable operation characteristics and perspectives of the large-current HTS STARS conductor

The High-Temperature Superconducting (HTS) magnet option has been explored for fusion reactors as well as for next-generation fusion experimental devices. The Stacked Tapes Assembled in Rigid Structure (STARS) conductor uses HTS tapes with simple stacking without twisting and transposition. A practically applicable STARS conductor is presently being developed with an operating current of 18 kA at 20 K temperature and ~15 T magnetic field. The conductor is required to have a high current density of 80 A/mm2. For the second stage of the conductor development, internal electrical insulation is applied between the copper stabilizer casing and the outer stainless-steel jacket, and a 6-m conductor sample was fabricated in a solenoid coil shape with a 600-mm diameter and three turns. The coiled sample was tested in 8 T, 20 K using a facility equipped with a maximum 13-T, 700-mm bore solenoid coil. Excitation up to the rated current of 18 kA was successfully attained with stable operation. The characteristics of the conductor observed during the excitation test are discussed.

Electrical diagnostics for high voltage tests in MITICA

ITER is an experimental fusion device and requires two Neutral Beam Injectors (NBIs) for plasma heating and current drive which have to supply a power of 16.5 MW each, delivered by a 1MeV deuterium beam. This D0 beam is generated starting from negative ions, produced inside a RF driven ion source, and accelerated by a Multi-Aperture Multi-Grid electrostatic accelerator insulated in vacuum.MITICA (Megavolt ITER Injector & Concept Advancement) is a full-scale prototype of the ITER Heating Neutral Beams (HNBs), devoted to test and optimize the design, under construction in Padova at the Neutral Beam Test Facility (NBTF) hosted by the Consorzio RFX. One of the most critical requirements in MITICA concerns the capability to withstand the nominal voltage of 1MV on a single long gap (≈1m), which separates the beam source and the vacuum vessel, insulated by vacuum or low pressure gas (10−6-10−3 mbar range) and with large size electrodes. In order to address this issue, a dedicated experimental test campaign exploiting a 1:1 mockup of the MITICA Beam Source will be carried out at NBTF before the installation of the actual Beam Source.During this test campaign, a set of electrical diagnostics will be installed to identify the location of the breakdowns as well as to measure the magnitude of the microdischarge currents (the precursors of vacuum breakdowns), and arc currents to provide useful information to optimize the electrostatic design of the ITER HNB and to validate the numerical models. This paper describes the development and the design of these diagnostics system based on electrical measurements.

Combined Capacitor–Resistor Energy Transfer System to Increase Plasma Current in RFX-Mod2

RFX-mod is an experimental fusion device, which contributes to plasma physics studies both in reversed field pinch (RFP) and tokamak configurations. Its high flexibility, due to an active magnetohydrodynamic (MHD) control system and the modular coils power supply (CPS) system, allowed operating RFX-mod in a wide range of experimental conditions with a plasma current up to 2 MA. Experiments with such high plasma current allowed the study on new promising confinement regimes, dominated by a self-organization process with the generation of a helical structure in the plasma core. Presently, RFX-mod is under a significant upgrade (RFX-mod2) to extend the operational scenarios increasing the proximity between the conductive shell and the plasma. This main modification of the magnetic front-end, together with other main improvements, is expected to increase the performance of the machine in both magnetic configurations. In the frame of the studies to exploit the new potential of RFX-mod2 achieving a higher plasma current and longer flat-top duration, a solution based on additional magnetic energy storage has already been proposed. The study of RFP physics at higher plasma current could be crucial to confirm positive trends, such as electron temperature and persistence of quasi single helicity states with the plasma current, and to explore and achieve improved confinement states. This article presents an alternative reconfiguration of the poloidal power supply system of RFX-mod2 based on a combined resistor–capacitor energy transfer system. This system allows to store energy in capacitor banks during the first phase of the plasma current ramp-up and release it to the plasma when the magnetizing current changes polarity, driving the plasma current over 2.5 MA. The proposed upgrade does not involve radical poloidal CPS modifications, maintaining the present converters number and ratings and remaining within the power limits of the main power transformers (300 MVA).

Multi-scale turbulence simulation suggesting improvement of electron heated plasma confinement

Turbulent transport is a key physics process for confining magnetic fusion plasma. Recent theoretical and experimental studies of existing fusion experimental devices revealed the existence of cross-scale interactions between small (electron)-scale and large (ion)-scale turbulence. Since conventional turbulent transport modelling lacks cross-scale interactions, it should be clarified whether cross-scale interactions are needed to be considered in future experiments on burning plasma, whose high electron temperature is sustained with fusion-born alpha particle heating. Here, we present supercomputer simulations showing that electron-scale turbulence in high electron temperature plasma can affect the turbulent transport of not only electrons but also fuels and ash. Electron-scale turbulence disturbs the trajectories of resonant electrons responsible for ion-scale micro-instability and suppresses large-scale turbulent fluctuations. Simultaneously, ion-scale turbulent eddies also suppress electron-scale turbulence. These results indicate a mutually exclusive nature of turbulence with disparate scales. We demonstrate the possibility of reduced heat flux via cross-scale interactions.

Microstructure and Thermal Property of Inert Plasma-Sprayed Tungsten Coating on SiC-Coated Carbon/Carbon Composites

The coating technology of tungsten on carbon/carbon (C/C) composite is an important issue for fusion experimental device components. In this study, an interlayer of chemical vapor deposition SiC between tungsten coating and C/C substrate was used. A tungsten coating 320 μm thick was successfully deposited on SiC-coated C/C substrate by inert plasma spray. The microstructure, roughness, and constituents of W-SiC-C/C composite materials were investigated using a scanning electron microscope, energy dispersive X-ray spectroscope, X-ray diffractometer, and atomic force microscope. The tungsten coating structure that may prevent crack propagation essentially consisted of a stacked lamellar columnar microstructure and particle cluster microstructures. The interfaces between the tungsten and SiC coating and between the SiC coating and the C/C were clear. The SiC interlayer acts as a barrier for carbon and tungsten diffusion. The thermal conductivity of the system was calculated by the mixture rule, which was 47.33 to 82.35 W/(m·K). The thermal expansion coefficient of W-SiC-C/C was negative at room temperature and up to 1.5 × 10−6/K for elevated temperature.

Spontaneous dissociation of H2 and high energy barrier of H invasion on W doped Al2O3(0001) surface

In view of the wide use of tungsten in fusion experimental devices and the importance of hydrogen isotopes permeation, here we studied the adsorption, dissociation, diffusion and invasion behavior of hydrogen on W doped α-Al2O3 (0001) surface. Based on the first-principle approaches, we found the W substitution for a top surface Al atom is the most energetically favorable. H2 molecule prefers to be adsorbed on the surface W and spontaneously dissociates into two H anions. Near the W defects, H atoms favor to be adsorbed at the W and Al sites rather than O sites on the surface, and within the subsurface layer H can only bond to W stably. As a result, H migration to subsurface should occur around W with an energy barrier as large as 4.22 eV which is much larger than the 1.91 eV around the O atom on undoped α-Al2O3 (0001) surface. These findings suggest that W surface doping is beneficial to α-Al2O3 as tritium permeation barrier.

Automated signal classification in the C-2W fusion experiment.

In TAE Technologies' current experimental fusion device, C-2W (also called "Norman"), record breaking, advanced beam-driven field-reversed configuration plasmas are produced and sustained in steady state utilizing variable-energy neutral beams, expander divertors, end-bias electrodes, and an active plasma control system. With a rapid shot-pace and an extensive number of data channels, the amount of data generated necessitates automated signal processing. To this end, a machine learning algorithm consisting of a multi-layered neural network as well as other data processing software has been designed for signal feature identification, allowing for accurate and fast signal classification, anomalous condition detection, and providing for signal pre-processing. With a small set of training data, the neural network can be "bootstrapped" to provide a robust classification system while minimizing human oversight. An example using data from the theta pinch plasma formation pulsed power system is presented. With an overall accuracy of ∼97%-having classified more than 5 × 106 pulsed power signals-the classification scheme is more than sufficient to fine-tune machine set points. However, this technique can be used for near-real-time preprocessing of any plasma physics signal and has wide ranging application in fusion experiments for the varied data produced by plasma diagnostics.

Improvement of Ic degradation of HTS Conductor (FAIR Conductor) and FAIR Coil Structure for Fusion Device

As a high-temperature superconducting (HTS) conductor with a large current capacity applicable to a nuclear fusion experimental device, REBCO (REBa <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">2</sub> CuO <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">y</sub> ) tapes and high-purity aluminum sheets are alternately laminated, placed in a groove of an aluminum alloy jacket having a circular cross section, and the lid is friction-stir welded. To make the current distribution and mechanical characteristics uniform, the conductor is twisted at the end of the manufacturing process. In the early prototype conductor, when the I <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">c</sub> was measured in liquid nitrogen under self-magnetic field conditions, I <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">c</sub> degradations were observed from the beginning, and the characteristic difference between the two prototype samples under the same manufacturing conditions were large. Furthermore, I <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">c</sub> degradation was progressed by repeating the thermal cycle from room temperature to liquid nitrogen temperature. This I <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">c</sub> degradation did not occur uniformly in the longitudinal direction of the conductor but was caused by local I <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">c</sub> degradation occurring at multiple locations. If the conductor was not manufactured uniformly in the longitudinal direction, the difference in thermal shrinkage between the REBCO tape and the aluminum alloy jacket caused local stress concentration in the REBCO tape and buckling occurred. Element experiments to explain this mechanism were conducted to clarify the conditions under which I <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">c</sub> degradation due to buckling occurs. Then prototype conductors were tested with improved manufacturing methods, and as a result, I <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">c</sub> degradation could be suppressed to 20% or less. We have achieved the prospect of producing a conductor with uniform characteristics in the longitudinal direction.

Availability and Reliability Assessment for ITER Radial X-Ray Camera System in Final Design

ITER is an international experimental nuclear fusion device. In the project, the RAMI approach (reliability, availability, maintainability, and inspectability) has been adopted for technical risk control to mitigate all the possible failures of components in preparation for operation and maintenance. The availability and reliability assessment is required and is one of main design contents in the ITER radial X-ray camera diagnostic (RXC) system final design review phase. The availability and reliability assessment, based on the RAMI program, ensured the system with a very high performance to measure the X-ray emission and research the MHD of plasma with high accuracy on the ITER machine. A functional breakdown was prepared in a bottom-up approach, and the system was divided into three main functions, eight intermediate functions, and 31 basic functions which are described using the IDEFO method. Reliability block diagrams (RBDs) were prepared to calculate the reliability and availability of each function under assumption of operation conditions and failure data. Initial and expected scenarios were analyzed to define risk-mitigation actions. The initial availability of RXC system was 86.13%, while after optimizations were taken the expected availability was 92.57%. The initial reliability for all functions of ITER RXC system was 92.3% without spares and 94.7% with spares over 16 months. A failure modes, effects, and criticality analysis (FMECA) was performed to the system initial risk. Criticality charts highlight the risks of the different failure modes with regard to the probability of their occurrence and impact on operations. There are 29 risks for the initial state, including 16 major risks. No major risk remains after taking into account all the actions. It was assessed that the RAMI analysis results meet the project requirement in the final design phase and the results will be qualified further with the system manufacture going ahead.

Analyses of deuterium retention in tungsten and graphite first wall materials by laser-induced ablation spectroscopy on EAST

• Combined LIBS and LIAS measurements were applied to determine the dynamic deuterium retention in C and W during EAST discharges. • LIAS measurements show more than three times higher deuterium retention in graphite than in tungsten during EAST discharges. • The deuterium retention on the graphite PFC is saturated within 0.2 s plasma exposure for limited magnetic configuration when the plasma actively touches the first wall. • The deuterium retention in the start-up phase is significantly larger than during the discharge for diverted magnetic configuration. Plasma-wall interactions at the first wall in fusion experimental devices are critical for the life time of plasma-facing components and tokamak operation. In-situ measurements of these plasma-wall interaction processes are required in long pulse devices in order to understand better the underlying mechanisms as well as monitoring of the tritium content of the first wall as an immediate safety application. In this work we present measurements of the deuterium retention in the first wall of EAST using laser-induced ablation spectroscopy. The diagnostic is applied during plasma operation and the dynamic retention for graphite and tungsten as plasma-facing materials are compared, showing more than three times higher short-term retention in graphite than in tungsten. Laser-based analysis is applied during a shift of the vertical positions of the plasma column in limited as well as in the start-up phase of discharges in diverted magnetic configuration. Plasma core densities of n e = (1.5–3.2) × 10 19 /m 3 are chosen to vary the ion flux to the first wall and study saturation effects of deuterium on the surface first wall materials during tokamak plasma operation. A linear dependence of the retention from ion fluence is observed for graphite as PFC up to 3 × 10 18 D + /cm 2 , followed by a constant retention up to fluences of 3.5 × 10 19 D + /cm 2 for nearby first wall temperatures of T PFC ≈ 350 K.

Development of a real-time framework for parallel data stream processing⋆

Abstract This paper presents the Korean Superconducting Tokamak Advanced Research (KSTAR) real-time framework for the parallel data stream processing framework (RT-ParaPro). RT-ParaPro is a framework used to develop a program that simultaneously processes data streamed over a real-time network, sends the data over a network, and archives the data in real-time. In most fusion experimental devices, each device processes the data needed for real-time control and transmits them to the other real-time systems in real-time via the network. By using RT-ParaPro, it is possible to simplify the configuration of a program that performs a series of processes and shorten the development time. Unlike other real-time frameworks that focus on real-time control, RT-ParaPro is specialized for the parallel data stream processing and transmission of data over a real-time network. By using this framework, the L-H transition detection system using machine learning (LHML), which determines whether plasma is in low-confinement mode (L-mode) or high-confinement mode (H-mode) in real-time by using machine learning, and the reflective memory (RFM) archiving system, which stores various RFM channel data to MDSplus, has been developed and operated in KSTAR. To evaluate the real-time performance of this framework, we tested the consistency of the thread period by varying the period of the thread. The test results show that the thread control period is consistent. The period of the thread has a jitter of about 8 μ s not only at a low control cycle rate (1 kHz) but also at a control cycle rate of 100 kHz.

Topology optimization for superconducting magnet system in helical fusion reactor

In magnetic-confinement-type fusion reactors, the strong electromagnetic force generated by the superconducting coil is of the order of several tens of MN/m. To support this huge force, a strong and heavy coil support structure is required. The total weight of each magnet designed for conventional fusion reactors or experimental devices is several times greater than the ideal value. In recent years, structural optimization has advanced, owing to improvements in computer-aided design and finite-element analysis. For instance, the topology optimization method contributes novel designs that overturn common sense. In the present study, topology optimization is applied to the design of the magnet system used in a helical fusion reactor to reduce the weight of the magnet while maintaining the mechanical soundness of the component. A weight reduction of more than 25% from the conventional design could be achieved.