Journal of Fusion Energy最新文献

The supercritical carbon dioxide (sCO₂) cooled Lithium–Lead (LiPb) dual function blanket is an advanced blanket concept design proposed by Chinese Fusion Engineering and Test Reactor (CFETR), which has the characteristics of high inherent safety and high thermoelectric conversion efficiency. In order to study the operational safety characteristics of the fusion reactor blanket Auxiliary Cooling System (ACS) and ensure the safe and stable operation of the system, this paper establishes a dynamic simulation model of key equipment for the fusion reactor blanket ACS, and develops a system analysis program. Based on the self-programming system program, this paper analyzes the thermal safety effects of various heating powers and flow rates on the sCO₂/LiPb dual function blanket, and conducts research on the operational safety characteristics of the system under expected shutdown transient events and varying degrees of Unprotected Loss of Heat Sink (ULOHS) accidents. The results indicate that sCO₂ and LiPb can effectively remove plasma radiation heat flux and nuclear thermal power, ensuring the thermal safety of the blanket. When the LiPb mass flow rate is 60% of the rated value, the outlet temperature of LiPb in the breeding zone exceeds its design limit of 1273.15K. When the LiPb flow rate is 30% of the rated value, the temperature of the cooling plate material in the blanket exceeds its design limit of 798.15K. Under expected shutdown operating conditions, the system can effectively export residual heat from the blanket. The ultimate operating condition of ULOHS accident is a 95% loss of sCO₂ flow on the secondary side, and after 3000 s, the LiPb temperature inside the blanket exceeds the limit value. This study can provide technical support for system regulation and operation control.

We report the implementation and technical evaluation of a shot-by-shot data transfer system from the GAMMA 10/PDX tandem mirror device to the LHD LABCOM system via SNET, as part of the Fusion Virtual Laboratory project. Since 2008, we have transferred approximately 1.5 TB of experimental data annually. In 2023, we successfully established real-time CAMAC data transfer on a per-shot basis. This paper details the network architecture, data acquisition systems, transfer protocols, and operational reliability. The system enables remote access and supports collaborative research by providing timely and structured data to the LABCOM database. The technical performance and future development plans are also discussed.

The atomic data of heavy elements, especially rare-earth metals, plays a crucial role in enhancing our understanding and interpreting kilonova spectra and underlying astrophysical processes. Among these elements, Erbium (Er) is particularly intriguing because it is important for opacities of the kilonova observed in 2017 (GW170817). In order to assess the atomic data, optical spectra of Er ions were precisely measured in (385-400,hbox {nm}) at Large Helical Device (LHD). In the present experiment, Er was injected into the core plasma of LHD through carbon pellets containing Er powders. The electron density and temperature of the Er-contained C pellet ablation cloud were obtained to be (1.6 times 10^{22},hbox {m}^{-3}) and 1.4 eV using the Stark broadening of a C II line and the Boltzmann plot of Er II lines, respectively. Transition probabilities of observed Er II lines were assessed using the Boltzmann plot analysis. Recent measurements with laser-induced breakdown spectroscopy (LIBS) of an Er II line at 393.86 nm were confirmed by the present work.

Fusion energy is regarded as a promising energy source due to its abundant fuel reserves, high energy efficiency, and reduced radioactive waste compared to fission. However, risks such as radioactive leakage, extreme operational conditions, and hazardous materials (e.g., cryogens, magnets) necessitate thorough safety assessments. Some countries have begun to develop their own fusion experimental devices to gain an advantage in the future energy distribution. Although fusion reactors have their inherent safety, there is still a risk of radioactive leakage that threatens the environment and people around, so it is necessary to perform a risk assessment of fusion devices. The probabilistic safety assessment (PSA) technology currently used in commercial pressurized water reactors is not applicable to fusion devices at this stage due to their structural particularity. To address this issue, this paper analyzes the differences between a fusion device and a fission device from three aspects: structural design, radioactive source term, and safety system, and proposed a preliminary probabilistic safety assessment method for fusion devices, with a case study on the ITER facility, which has been applied to a fusion experimental device under design and achieved good analytical results.Note that this method is currently validated only for ITER-like experimental devices, and its extension to other fusion plants requires further scenario-specific adjustments.In addition, the case study on ITER is based on historical design data and serves as an example application of this framework. The result does not represent the current security assessment of ITER.

We review global gyrokinetic simulation studies on plasma transport in the Large Helical Device using XGC-S. XGC-S is an extended version of X-point Gyrokinetic Code for stellarators and has been progressively verified throughout the code development process. Verification tests of neoclassical transport successfully demonstrate the generation of an ambipolar electric field due to ripple-trapped particles. We perform quasi-linear analyses of the ion temperature gradient mode under the influence of the ambipolar electric field. The results reveal that the ambipolar electric field and the heavy hydrogen component in mixed isotope plasmas can lead to the favorable isotope effect observed in recent deuterium experiments. We also present recent efforts in code development toward whole-volume simulations, including the helical divertor region. A mesh generation scheme based on field-line tracing and the construction of curved surfaces perpendicular to the magnetic field would be promising for global field calculations in the whole-volume simulations.

This study has investigated the effect of particle transport of fuel ions, deuterium (D) and tritium (T), and helium ash (He) as well as their heat transport on the nuclear fusion output in a DEMO reactor. The result was assessed and possible start-up scenarios for DEMO was discussed. In the study, the integrated tokamak modeling code, TASK/TR is used to investigate the impact of the neutral beam injection (NBI) fuel source on the plasma density distribution. It was found that, due to the dilution from burn-up and helium ash, the fuel density profile becomes either flat or hollow in the core region when fuel is supplied only from the periphery. The reduction of core fuel dilution by NBI is discussed in the context of the results, highlighting the potential of NBI to mitigate the hollow density profile and enhance the central fueling during startup.

The impact of edge magnetic islands on divertor detachment in LHD is investigated, with emphasis on thermal instability and thermal equilibrium. During a density ramp-up, as the edge plasma temperature is reduced, radiation is enhanced in cases with an edge magnetic island compared to those without one, and the detached plasma state remains stable. In contrast, in the absence of the island, the increase in radiation becomes uncontrollable, ultimately leading to a radiation collapse of plasma. Analysis of thermal instabilities indicates that the X- and O-points of the island are particularly susceptible to thermal instability due to their distinct magnetic topologies. Impurity radiation measurements reveal that, during density ramp-up, radiation initially emerges around the island’s X-point. Following the detachment transition, the location of peak radiation shifts to the O-point, where signatures of volume recombination are observed. Numerical simulations of edge plasma transport reproduce these dynamic trends, which reinforces the experimental interpretation. Thermal instability growth rates estimated from experimental data indicate that, within the island’s O-point, the growth rates decrease as detachment deepens; in contrast, in the absence of the island, the growth rates continue to increase, approaching radiation collapse. Additional aspects of thermal instability and radial thermal equilibrium are discussed to elucidate the factors contributing to detachment stabilization and to outline remaining challenges for future investigations.

Long pulse discharge experiments have been conducted at LHD as one of the challenges in the quest for a fusion reactor. The plasma could be sustained for 48 min, using 1.2 MW of heating power from the ion cyclotron range of frequencies (ICRF) and electron cyclotron heating (ECH) power. The line-averaged electron density was 1.2 × 10¹⁹m^-3, the ion and electron temperature was 2 keV, and the input energy reached 3.36 GJ. The ICRF antennas used were the HAS (handshake form) and the FAIT (field-aligned-impedance-transforming) type, which upgraded and replaced the existing PA (poloidal array) antennas, in addition to the existing PA antenna. These antennas had the following characteristics. The PA one had a Faraday shield on one antenna strap removed. The HAS antenna’s straps were aligned in the toroidal direction and was phase controllable in the toroidal direction. The FAIT antenna had a built-in impedance transformer to obtain high plasma loading resistance. Long pulse discharges are often terminated by an influx of impurities, which are the result of exfoliation of deposits left during the discharges. Overcoming this impurity problem is one of the challenges of a steady state operation.

At the National Institute of Fusion Science (NIFS), the deuterium plasma experiment was conducted using the large helical device (LHD) from 2017 to 2022 for the high performance of the plasma experiment. Through this experiment, a small amount of tritium was produced by D-D fusion reaction and released to the atmospheric environment through the stack. To understand the impact of tritium on the environment, environmental tritium monitoring was conducted before, during, and after the experiment for public acceptance and in accordance with local governments. From this monitoring, no impacts were observed on monthly precipitation and pine needle samples at the NIFS site. As the result of a comprehensive assessment combined with atmosphere and environmental water monitoring, it was concluded that the impact of discharged tritium from the stack of LHD to the surrounding environment would be none and/or negligibly small.

The Impurity Powder Dropper (IPD) is a device capable of injecting controlled amounts of sub-millimetre powder into the plasma under the action of gravity. In 2019 the IPD was first installed on the Large Helical Device (LHD) in Japan, with the aim of improving the plasma performances through real time boronization and assessing the compatibility of this technique with steady state operation. Extensive series of experiments have been performed using the IPD, focused on the improvement of the plasma performance via low-Z powder injection and the understanding of the underlying physical phenomena. In this article, we review the experiments that took place in the period 2019-2024. The main results include the demonstration of the improvement of the wall conditions (reduction of intrinsic impurity content, wall recycling) both on a shot-to-shot basis and in real time. Furthermore, a reduced-turbulence improved confinement regime has been observed coincident with powder injection, resulting in an increase of the plasma temperature of the order of 25%, with enhancements that can reach up to 50% for ion temperature.