Journal of Fusion Energy最新文献

As fusion energy technologies approach demonstration and commercial deployment, understanding public perspectives on future fusion facilities will be critical for achieving social license. In a departure from the ‘decide-announce-defend’ approach typically used to site energy infrastructure, we develop a participatory design methodology for collaboratively designing fusion energy facilities with prospective host communities. We present here our findings from a participatory design workshop that brought together 22 community participants and 34 engineering students. Analysis of the textual and visual data from this workshop shows a range of design values and decision-making criteria with ‘integrity’ and ‘respect’ ranking highest among values and ‘economic benefits’ and ‘environmental protection/safety’ ranking highest among decision-making criteria. Salient design themes that emerge across the facility concepts include connecting the history and legacy of the community to the design of the facility, respect for nature, care for workers, transparency and access to the facility, and health and safety of the host community. Participants reported predominantly positive sentiments, expressing joy and surprise as the workshop progressed from learning about fusion to designing the hypothetical facility. Our findings suggest that carrying out participatory design in the early stages of technology development can invite and make concrete perspectives on public hopes and concerns, improve understanding of, and curiosity about, an emerging technology, build toward social license and inform context-specific development of future fusion energy facilities. Drawing on our findings and design process, we propose a prototype playbook for participatory design that will be developed further in future research. We recommend that fusion development teams (1) consider using participatory design approaches at multiple junctures throughout the fusion technology or facility development process, (2) build capacity to carry out such participatory engagements and (3) design standardized but adaptable technologies and facilities. We invite fusion technology developers to use and adapt our playbook for their own projects.

Graphical Abstract

A study of aneutronic fusion, which refers to fusion reactions that do not produce neutrons, has been conducted on the Large Helical Device (LHD). For the D–³He study, feasibility investigations of detecting charged fusion products have been performed. High-energy neutral beams, lost proton detectors, and gamma-ray detectors were planned for installation to validate the D–³He fusion reactions. Numerical calculations show that the expected D–³He fusion rate is 2.7 × 10¹⁶ s^-1 with most of the protons being lost immediately from the plasma. For the p–¹¹B study, the total alpha particle emission rate was estimated to be 10¹⁴ s^-1, and the loss points of alpha particle distribution were calculated. Based on these numerical simulations, we installed an alpha particle detector on a manipulator and positioned it at the bottom of the LHD. The time evolution of the alpha particle detection rate, measured by the detector, was found to be consistent with the predictions from the numerical simulations, demonstrating the first successful observation of the p–¹¹B reaction in a magnetic confinement system.

Thomson scattering diagnostic systems are widely used to measure the local electron temperature and density of plasmas, which are one of the most important plasma parameters. The Large Helical Device (LHD) Thomson scattering system was designed and developed from 1991 to 1998, and has been in operation without any major problem since the LHD second experiment campaign in 1998. The LHD Thomson scattering system measures the pseudo-continuous time evolution of electron temperature and density profiles at 144 spatial points along the LHD major radius. The LHD Thomson scattering system can measure electron temperature and density spatially from the inner boundary to the outer boundary, and temporally from the birth to the destruction of LHD plasmas. The performance is still one of the best in the world. In this paper, we discuss improvements and newly obtained results of this system with emphasis on the results from 2010 to 2024.

To study microscale turbulence, two non-invasive scattering instruments that use electromagnetic waves in the microwave to millimeter-wave range have been installed at the Large Helical Device. One instrument is a Doppler reflectometer, which is suitable for observing turbulence with relatively low wavenumbers. Three circuit systems were constructed for this instrument. The Doppler reflectometer allows a very large number of spatial points (more than 30) to be observed simultaneously in the radial direction and toroidal correlation analysis to be conducted. The other instrument is a two-frequency millimeter-wave scattering system, which was developed to observe turbulence at relatively high wavenumbers. This scattering system has multiple antennas in a vacuum vessel. It can be used to study turbulence anisotropy or, in combination with the Doppler reflectometer, the response of turbulence at various scales.

Against the backdrop of the accelerated advancement of the International Thermonuclear Experimental Reactor project, the Central Solenoid Model Coil (CSMC) of the China Fusion Engineering Test Reactor (CFETR) plays a pivotal role in achieving efficient plasma confinement and control. During the assembly of CSMC, inevitable assembly errors can disrupt both the central symmetry around the central axis and the planar symmetry with respect to the mid-plane of the Nb₃Sn and NbTi coils. This symmetry disruption leads to the generation of substantial asymmetric radial and axial offset electromagnetic forces within the Nb₃Sn and NbTi coils, which potentially jeopardize the preload stability and structural integrity of supporting components. To systematically investigate the stability of CSMC under the influence of assembly errors, this study first established a calculation model based on electromagnetic field theory for three typical assembly error scenarios, followed by the computation of corresponding offset electromagnetic forces. Second, the linear instability analysis method was utilized to determine the instability modes of CSMC under different offset electromagnetic force conditions. Finally, an extreme offset state was selected for nonlinear instability analysis to quantify the critical load triggering instability. The research results indicate that, under extreme assembly errors, preload rods emerge as the weakest structural components susceptible to instability. Notably, the critical instability load demonstrates an 8.3-fold margin over the combined operational loads. This study offers critical data references for optimizing the assembly precision and conducting safety margin evaluations in CSMC operation.

Fast-ion transport driven by Alfvén eigenmodes (AEs) is one critical issue facing fast-ion confinement in magnetic fusion device. In the DIII-D tokamak experiment, stiff transport of fast-ions increased with increasing neutral beam (NB) injection power when the amplitudes of multiple interacting AEs exceeded a certain threshold. These experiment results are supported by simulation studies that predict monotonically degrading fast-ion confinement and profile stiffness with increasing beam power. To investigate the universality of the fast-ion profile stiffness dependence on AE amplitude, an experiment was performed at the Large Helical Device (LHD) to scan the injection current of the NB and vary the AE amplitude. Under the experimental conditions, the AE amplitude increased linearly with NB injection power. The red shifted FIDA intensity between 663 and 665 nm, corresponding to the energy range of 98–166 keV in the ctr-direction, was used for estimating the radial profile of the fast-ion density. Evidence suggests stiffening of the fast-ion profile and degraded confinement, corroborated by a reduced neutron emission rate compared to simulations. This is consistent with the experimentally observed reduction in the expected neutron emission rate. We have demonstrated that under AE-prone confinement conditions, even if the fast-ion source increases due to NB injection, they experience enhanced transport by AEs and do not increase in density.

This paper presents the first studies of combined glow + microwave discharges at the TOMAS facility, in a volume of ~ 1.1 m³, and discusses microwave propagation in the plasma. The combined discharge was realized by injection of additional microwave power (0.4–1.5 kW) at 2.46 GHz into the argon plasma of the glow discharge. The studies have shown that the injection of additional microwave power allows to reduce the voltage on the glow discharge. The maximum observed voltage decreases in the combined glow + microwave discharges compared with a reference glow discharge at ≈ 220 V. The dependence between the voltage decreases and the injected microwave power is linear. This effect of the combined glow + microwave discharges provides flexibility to study particular aspects of wall conditioning techniques relevant to larger devices.

The Resistive Wall Mode (RWM) is frequently linked to external kink (XK) instability, which arises from high-pressure gradients in fusion devices. This study emphasises the critical importance of adjusting the phase of Resonant Magnetic Perturbation (RMP) coils to suppress the RWM instability, while considering factors such as plasma elongation and triangularity. Using the MARS-F code [Liu et al. 2000, Phys. Plasmas 7, 3681], we investigated how phase modulation of RMP coils affects the growth rates of RWM across various plasma profiles and thin-wall conditions. Our results demonstrate the effectiveness of phase modulation of RMP coils across different configurations, including elongation, triangularity, and normalised beta. Additionally, we found that toroidal rotation is crucial in suppressing RWM growth rates. These findings provide valuable insights for designing advanced high-confinement scenarios during tokamak operations.

In high-temperature plasma experiments using deuterium gas in large fusion devices, tritium is produced by the deuterium fusion reaction. Although the amount of tritium produced is not large, it is a radioactive material, so it is essential to develop safe handling systems and obtain public acceptance. Here, we summarize tritium safety management and tritium behavior in the facility, and have monitored the results of six years of deuterium plasma experiments in the Large Helical Device (LHD). More than 95% of the tritium exhausted from the LHD vacuum vessel was recovered by a tritium removal system. The extremely low concentrations of tritium discharged from the stack into the environment were monitored for each chemical form of tritiated water vapor, tritiated molecular hydrogen, and tritiated hydrocarbons, and were verified to be well below the levels specified by management at the National Institute for Fusion Science (NIFS). Although the tritium handled in the LHD deuterium experiment was a small amount, the operational experience and the instruments thereby developed would likely be useful for tritium safety management in future fusion reactors.

The primary purpose of the ITER Erosion Deposition Monitor (EDM) is to track the erosion and deposition conditions of the Divertor Vertical Targets, as well as to monitor any changes in topology and surface damage resulting from plasma-wall interactions at these targets. The optical box of this diagnostic system, which is mounted on one of the Divertor Cassettes beneath the Divertor Dome, serves to provide a rigid support and protective shielding for the optical components housed inside. The EDM is an optical diagnostic system, whose accuracy depends on the integrity of the mechanical parts. The most critical in-vessel parts of this diagnostic system can be found on one of the Divertor Cassettes of the ITER machines, exposed to high heat and electromagnetic (EM) loads. The design challenges involve finding a solution for the proper heat transfer between the different components and besides, it is creating a solid design that can withstand the loads the optical box is exposed to. To achieve the required design level, the optical box and its components went through several design changes and simulation stages. This paper focuses on the challenges of the mechanical development of the parts that can be found on the Divertor Cassette. These components are the structural elements of the optical box and the support brackets of the mirrors inside the box. The paper presents the results of the analyses and the mechanical solutions of the structure of the box and the mirror supports which could withstand the high heat load and mechanical stresses coming from the EM loads.