Journal of Fusion Energy最新文献

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.

Divertor plasma-facing components (PFCs) in a tokamak are typically designed to withstand average steady-state heat loads of about 5–10 MW/m², a limit that applies to both solid and liquid lithium (LL) PFCs. Exceeding these design values can result in surface damage to tungsten PFCs or excessive lithium (Li) evaporation in liquid lithium divertor (LLD) PFCs. Since exceeding the divertor heat load limits has serious consequences, it is therefore prudent to develop a tool to reduce the divertor heat load and bring the heat load to within the design limit without affecting the plasma performance. Active low Z impurity injection such as Li has been suggested as a potential solution to mitigate excess heat flux as suggested previously, given that non-coronal radiation can be quite large ~ 20–30 MJ per mole of injected Li. Li is considered desirable for reducing the edge neutral recycling helping to improve plasma energy confinement. In this paper, we model the Impurity Power Dropper (IPD) to investigate its potential of divertor heat flux control. The IPD is typically located at the top of the tokamak device and uses a vertical drift tube of a few meters. In the 2 m drift tube case, the IPD powder is accelerated to ~ 6 m/sec before reaching the plasma with the upper divertor configuration, matching the condition for the in-board side pellet injection case. By modeling the IPD geometry we determined the IPD powder deposition profile, and thus the non-coronal radiation and ionization profiles in time as well. From the enhanced radiation power loss, it is therefore possible to reduce the divertor heat load using the divertor simulation code. The IPD divertor heat flux control can be tested in the facilities with IPD including ST-40, DIII-D, EAST, WEST and NSTX-U.

This paper reviews studies of non-Maxwellian electron velocity distribution function (EVDF) measured via line emission spectroscopy and spectropolarimetry in the Large Helical Device (LHD). Information on where and under what conditions the non-Maxwellian EVDFs are generated can substantially affect plasma confinement. Since different atomic transitions exhibit different sensitivities to momenta of incident electrons, spectroscopic analysis of line intensities and their polarization enables the investigation of both the shape and anisotropy of the EVDF. Measurement techniques and their results are summarized across a broad temperature range, covering both edge and core plasmas in LHD. The results are compared with plasma parameters obtained from other diagnostic systems, and the dynamics of passing and trapped electrons are discussed.

Systems models and associated cost analyses are widely used within the fusion community to analyse tokamak designs, from prototype and demonstrator machines to potential commercial fusion power plants. To ensure the design programmes of fusion prototype/demonstrator power plants deliver a cost optimised design (within existing uncertainty limitations) the use of integrated cost modelling during the design process is essential. This integration produces holistic solutions in which engineering design choices and changes are directly represented in the cost results, allowing alternative solutions to be tested technologically and financially in the same analysis and cost estimates to be directly aligned with each specific design solution. Using examples from the STEP (Spherical Tokamak for Energy Production) programme this paper shows how implementing such an approach allows an interrogation of the design through the lens of cost-effectiveness, enabling a systematic exploration of potential trade-offs between performance and cost, highlighting cost drivers and interrogating the design aspects underpinning them, and facilitating holistic comparisons between design options. Including cost analysis into early design decisions through integrated cost modelling will drive a cost-optimised design; this is vital in proving that fusion power plants can be an economically viable energy source. One top-level example of this approach is understanding the critical size drivers and therefore cost drivers of the design, such as the inboard radial build for the STEP design. This understanding enables optimisation of this parameter within the relevant margins required to ensure performance (within design uncertainties).

This paper reviews the contributions of image observations to extending the duration of Ion Cyclotron Range of Frequencies (ICRF)-heated long-pulse discharges in the Large Helical Device (LHD). The plasma discharges were monitored using over 25 visible cameras, three fast-framing cameras, and various advanced plasma diagnostics, which revealed that most long-pulse discharges were interrupted by the following four events: termination of ICRF plasma heating due to arcing events in antennas, uncontrollable plasma density rise by outgassing from divertor plates, iron influx from plasma-facing components in the vacuum vessel, and carbon influx originating from the divertor regions. Image observations played a crucial role in mitigating the above four events that restricted the duration of long-pulse discharges by implementing appropriate countermeasures such as enhancing the cooling efficiency of the divertor plates, adopting new operational techniques to disperse the heat-load distribution, improving the ICRF antenna configurations, installing new additional ICRF antennas, and modifying the divertor configuration. Interruptions in long-pulse discharges were statistically analyzed using experimental data in three previous experimental campaigns, demonstrating a history of continuous efforts to extend the plasma discharge duration. This paper highlights the contributions of image observations over the past two decades, which have revealed inherent limitations in conventional magnetic plasma confinement devices that utilize carbon and iron plasma-facing components in sustaining steady-state plasma discharges. Knowledge obtained from statistical analysis provides valuable information for optimizing next-generation plasma confinement devices aiming at steady-state operation.

The EUV Imaging Spectrometer (EIS) on board the Hinode mission is capable of observing solar coronal plasma possibly in non-ionization-equilibrium. EUV emission lines from highly charged Fe ions observed in the solar corona are also produced in the Large Helical Device (LHD) and the compact electron beam ion trap (CoBIT). Time-dependent collisional-radiative model (CRM) for Fe ions is developed to diagnose those plasmas in the Sun and the laboratories by adopting the best available theoretical calculations of atomic parameters, as well as generating the experimental data.

The negative pressure ventilation system, serving as a critical subsystem within the tritium safety confinement system of the compact fusion energy experimental, is responsible for maintaining dynamic confinement functionality over the C2 and C3 confinement subzones during normal operational conditions, thereby restricting the leakage and dispersion of radioactive materials. Consequently, rational and reliable design of the tritium confinement negative pressure ventilation system is of paramount importance. Key aspects of the system process design include determining pipeline network dimensions and evaluating thermal-hydraulic characteristics. In this study, the system design requirements were established in accordance with the international standard ISO 16646:2024 and the operational specifications of the device. This involved calculating the system airflow capacity and proposing a preliminary process design scheme for the negative pressure ventilation system. A thermal-hydraulic computational model for the initial system configuration was developed using the fluid simulation software AFT Arrow. Steady-state simulations were conducted to predict the thermal-hydraulic behavior of the pipeline network under maximum ventilation load conditions, enabling the determination of critical equipment selection parameters. The findings of this study will lay the groundwork for subsequent research on negative pressure ventilation systems in compact fusion energy applications, providing essential technical insights for optimizing system performance and ensuring compliance with radiological safety standards.