Fusion Engineering and Design最新文献

The present work aims at defining a first design of the DEMO EU’s feeder thermal shield for the CFT, by exploring the temperature distribution resulting from radiation and conduction coming from different sources including the vacuum duct, the containment duct and the cryogenic He lines on the thermal shield. In order to achieve these results, at first a numerical FEM representation was produced and parameterized. To ease the burden of such a complex and highly non-linear solution, an analytical model however was created and validated against the numerical one with very good results. The sensitivity of several parameters was computed with respect to the temperature distribution of the thermal shield and the heat flow to the containment duct, finding a candidate configuration for the thermal shield. To complete the study, an evolutive optimization based on gradient descent was implemented to understand the influence of the circumferential positioning of the cooling He lines.

The stray magnetic fields generated by a plasma discharge impact the performance of neutral beam injection (NBI), leading to a decline in the plasma performance of the Korea Superconducting Tokamak Advanced Research (KSTAR). To evaluate the impact of the stray magnetic field on the NBI performance, a Monte Carlo simulation tool was developed. The simulation tool integrates the stray magnetic fields, NBI beam line components, and charge exchange processes comprehensively, allowing for quantitative analysis of the NBI performance reduction due to the stray magnetic fields. The high pressure of the beam chamber, along with the stray magnetic fields, causes a significant reduction of the beam power, emphasizing the need to maintain low vacuum pressure. However, the stray magnetic field does not significantly affect the injection angle of the beam particles reaching the tokamak. Predictive integrated simulations show that the decrease in beam performance due to stray magnetic fields can affect a degradation in plasma performance during long pulse discharge in KSTAR.

A data acquisition and control (DAC) system has been developed for the solid hydrogen pellet injection system. This injector is a gas gun type injector, where solid hydrogen pellet ice is formed using a closed cycle cryocooler, and high-pressure helium gas or a pneumatic punch is used to dislodge the pellet. The DAC system is based on National Instrument's embedded controller NI PXIe-8133, along with associated I/O DAC cards and the LabVIEW application. A graphical user interface (GUI) developed using LabVIEW software allows users to remotely control the pellet formation and injection process. The data is stored locally on a desktop computer or in cloud storage for further analysis. The images of the injected pellet were obtained by using a Phantom V-1210 high-speed camera at 100-Killo FPS, which reveals the size and speed of the pellet. The developed DAC system provides the flexibility needed to operate the injector remotely during the plasma discharge in a tokamak environment. The injector setup system has been successfully tested in a test bench operation, and it will be integrated with the ADITYA-U tokamak.

Tritiated water will be present in many demonstrator class fusion devices in a wide array of locations and concentrations. Accurate and fast measurement will be a key requirement to ensure suitable process monitoring, safety assurance and tritium tracking is in place. This work quantifies the deuterium detection limits achievable using a benchtop transmission Fourier transform infrared (FTIR) instrument and an industrialised attenuated total reflectance (ATR)-FTIR instrument to understand the capabilities of these techniques for real-time monitoring of aqueous tritium. Deuterium limits of detection of 4.62 × 10^-5 mol mL^-1 and 1.07 × 10^-3 mol mL^-1 were demonstrated with measurement times of 10s for the transmission FTIR and ATR-FTIR, respectively. These instruments are considered viable for the measurement of high concentration tritiated water and have many potential benefits associated with their deployment.

EAST (Experimental Advanced Superconducting Tokamak) has been in operation since 2006. With increasing operation time, the volume of plant data stored in MySQL has steadily grown to several billion rows. The current storage architecture centered around the relational database has shown poor performance when facing massive time series data, so it is crucial to adopt TSDB (Time Series Database) for storing plant data. In initial testing, IoTDB demonstrated a performance advantage of at least 2 times over other TSDBs in terms of both write throughput and large-scale queries for plant data management. However, as the inflexible underlying infrastructure of EAST, the plant database cannot be easily modified directly. To remedy this problem, we propose a MySQL-IoTDB Hierarchical Mechanism (MIHM). Specifically, we utilized a MySQL master–slave to IoTDB cluster design to seamlessly transfer the performance burden from relational database to TSDB IoTDB. Extensive tests on the EAST plant data demonstrate that MIHM-based plant data storage system has increased the write throughput by 20 times and the large-scale data querying speed by 100 times compared to previous systems.

The cryogenic vacuum system is an important subsystem of Comprehensive Research Facility for Fusion Technology Negative Neutral Beam Injector, which provides vacuum environment support for various experiments related to the beam generation and transport process. In the paper, a simulation and experimental study of the Comprehensive Research Facility for Fusion Reactor Negative Neutral Beam Injector (CRAFT NNBI) cryopump performance has been carried out. The Beam Line Vessel (BLV) and its components are modeled by Solidworks, the pressure distribution, gas molecular velocity distribution and adsorption uniformity of the crysorption pump in the BLV are analyzed by simulation using the Molflow software, at the same time, the performance of the south cryopump is tested experimentally using in-situ measurement method. The simulation results show that the average pressure near the Neutralizer is 8.5*10^-3 Pa, the average pressure near the Electrical Residual Ion Dump (ERID) is 8*10^-3 Pa, and the average pressure near the Calorimeter is 9*10^-3 Pa. The adsorption percentage fluctuates around 12%, which indicates that single structure plays the function of adsorption efficiently, the theoretical pumping speed of the south cryopump on H2 was calculated to be 7.34*10⁵ L/s, total theoretical pumping speed of two cryopumps can reach 3.67*10⁶ L/s. The experimental results showed that when the south cryopump was operation, the pressures near the Neutralizer were 2.8*10^-2 Pa, the pressures near the ERID were 2.7*10^-2 Pa, and the pressures near the Calorimeter were 7*10^-3 Pa, the actual pumping speed of the south cryopump on H2 was 6.37*10⁵ L/s, total actual pumping speed of two cryopumps can reach 3.1*10⁶ L/s.

To better study the magnetic fluctuations in the high-field side of EAST tokamak, a new high-field side magnetic probe array (HFS-MPA) has been developed on EAST recently. The HFS-MPA consists of 12 identical three-dimensional (3D) magnetic probes, which are mounted on the HFS wall with carefully designed arrangement. The HFS-MPA magnetic probes bring additional toroidal magnetic fluctuation measurements compared with the HFS regular magnetic probes which can only provide poloidal and radial magnetic fluctuation measurements. The upper limits of frequency and toroidal mode number (n) measurements of the magnetic fluctuations have been improved by comparing HFS-MPA with the HFS regular magnetic probes, i.e., 650 kHz vs 100 kHz and n = 23 vs n = 1. In developing the HFS-MPA diagnostic system, many practical challenges have been overcome and many special designs have been developed. These will be mentioned in the main subsystem description of the HFS-MPA diagnostic in this paper, which might be useful for developing new magnetic probe diagnostics in the future on EAST or other magnetically confined fusion devices. The calibration of the effective area and frequency response of the HFS-MPA is also described. The preliminary application in studying the frequency and propagation characteristics of the magnetic fluctuations with HFS-MPA compared with EAST regular magnetic probes shows that the HFS-MPA is well developed for plasma physics studies.

A statistical design of experiments was conducted to optimize a trimetallic catalyst formulation consisting of ruthenium, yttrium, and potassium on γ-Al₂O₃ (RuYK/ γ-Al₂O₃) for use as ammonia (NH₃) decomposition catalyst in the hydrogen isotope impurity processing for magnetically confined nuclear fusion systems. Optimal weight loadings of 6.9 wt-% Ru, 4.3 wt-% Y, and 12 wt-% K were determined through the design of experiments. The thermal stability of the catalyst was investigated through thermal cycling of the catalyst over 30 cycles. The optimized catalyst remained stable over the cycles under reducing conditions. As oxygen, carbon dioxide and water are the primary impurities in the Tokamak exhaust, the chemical stability of the catalyst was determined against these impurities. While these impurities initially decreased the NH₃ decomposition activity, the initial activity was attained once the impurity was removed from the stream.

This paper presents the results of experimental tests on samples made of copper coated with alumina layer, performed to assess the reliability of its dielectric properties for applications in the low temperature plasma at the edge of a fusion device. The cue of the study was related to the plasma facing components of the RFX-mod2 fusion device (Marrelli et al., 2019, Peruzzo et al., 2023, Peruzzo et al., 2019), devoted to the experimental study of the magnetic confinement of fusion plasmas in a variety of configurations, including the reversed-field pinch and the tokamak. In RFX-mod2 an in-vacuum copper shell for the passive stabilization of MHD modes will surround the plasma. To avoid potentially harmful electrical discharges, which could be induced by rapid transients of the plasma current, this structure must be covered with an electrically insulating layer. For RFX-mod2 an alumina coating was chosen, whose dielectric properties have been tested both in air and in the presence of weakly ionized plasma. Electrical tests, conducted on copper samples with alumina deposits of about $100\mu m$ thickness, revealed that the ceramic layer has a high electrical resistance value in air ($>1G\Omega$), but electrical discharges can occur in presence of a weakly ionized plasma, depending on compactness and porosity of the alumina layer, causing local melting of the alumina and expulsion of copper droplets from the substrate. Scanning Electron Microscope (SEM) analyses revealed that in the failed samples the ceramic layer was irregular and rough, with interconnected cavities and cracks, which could reduce its effective thickness and explain the dielectric breakdown at relatively low voltages ($<400V$). The analyses also showed that samples with a more compact layer present a higher dielectric strength in the presence of the plasma, highlighting that compactness and porosity play crucial roles in ensuring good insulation for materials in a plasma. This study led the definition of the requirements for the insulating coating of the plasma facing components of the RFX-mod2 fusion machine, however the results can be useful for other fusion and non-fusion plasma applications requiring electrical insulation, which can span from industrial devices to spacecrafts.

The fast-ion velocity distribution function is crucial for understanding fast-ion behavior and transport in future burning plasmas. However, direct measurements of this distribution are difficult due to its high-dimensional nature, necessitating inference from diagnostic data. To infer fast-ion velocity distributions in KSTAR experimental conditions, we explored the feasibility of using measurements from fast-ion D_α (FIDA) diagnostics. We assessed the reconstruction quality for two phantoms, representing a possible fast-ion distribution scenario and local velocity-space structures. We calculated the phase-space weight function of FIDA measurements, required for tomographic inversion, by modeling the measurements, and also developed a tomography code with Phillips–Tikhonov regularization. The phantom test results revealed limitations in the reconstruction capability of current FIDA systems in KSTAR, particularly near low-pitch regions. We also identified the influence of spatial bias of the weight function of the current FIDA systems. Introducing a new FIDA system to tomographic inversion process provided wider coverage in velocity space and the weight function with reduced spatial bias, thereby improving reconstruction capability, especially in low-pitch regions. We also scanned noise levels in the phantom tests and observed the benefits of using prior information to mitigate degradation of the reconstruction quality caused by noise.