Search Results (12)

High-temperature superconductor (HTS) magnet systems, especially those designed for fusion reactors, require effective and reliable monitoring to avoid damaging anomalies. In tokamaks, some of the magnetic coils are time-dependent, which causes strain and large inductive voltages within the magnet, rendering detection of incipient quench challenging. Ionizing radiation can also create material defects and lead to non-uniform degradation of conductors. The resulting decrease in critical current uniformity across the magnet, along with manufacturing defects, such as failure of structural materials or cooling systems, can all potentially initiate a quench. HTS magnets have a lower normal zone propagation velocity than low-temperature superconductors, and this causes normal zones to be localized, increasing the risk of permanent damage. Fiber optic sensors have several qualities that are essential in fusion systems. Unlike traditional voltage-based sensors, fiber optic cables are immune to the large electromagnetic fields present. This study presents and validates a fiber optic interrogation technique for monitoring magnetic confinement fusion and other high-temperature superconducting magnet systems. Coherent-phase optical time domain reflectometry (OTDR) allows for the high sampling rates (tens of kHz) necessary to quickly detect and mitigate quench events over the long distances required to monitor fusion magnet systems. This technique was demonstrated to successfully detect localized thermal transients at cryogenic temperatures as low as 6 K. These outcomes were also demonstrated using fibers embedded in HTS magnet coils at 77 K, verifying the potential for this interrogation technique’s use for failure detection in HTS coils. Full article

Tokamak technology, as the cornerstone of nuclear fusion energy, holds immense potential in achieving efficient plasma confinement and high energy densities. To comprehensively map the rapidly evolving landscape of this field, this study employs bibliometric analysis to systematically examine the research and development trends of tokamak technology from 2014 to 2024. The data are drawn from 7702 academic publications in the Scopus database, representing a global research effort. Additionally, the study incorporates 2299 tokamak-related patents from Google Patents over the same period, analyzing their technological trends to highlight the growing significance of tokamak devices. Using the R language and the Bibliometric package, the analysis explores research hotspots, institutional influences, and keyword evolution. The results reveal a multifaceted global landscape: China leads in publication output, and the United States maintains a leading role in citation impacts and technological innovation, with other notable contributions from Germany, Japan, South Korea, and various European countries. Patent trend analysis further reveals the rapid expansion of tokamak applications, particularly with significant innovations in high-temperature superconducting magnets and plasma control technologies. Nevertheless, the study identifies major challenges in the commercialization process, including plasma stability control, material durability, and the sustainability of long-term operations. To address these, the study proposes concrete future directions, emphasizing international collaboration and interdisciplinary integration. These efforts are crucial in accelerating tokamak commercialization, thereby providing a strategic roadmap for researchers, policymakers, and industry stakeholders to advance the global deployment of clean energy solutions. Full article

The Volumetric Neutron Source (VNS) is a pivotal facility proposed for advancing fusion nuclear technology, particularly for the qualification of breeding blanket systems, a key component of DEMO and future fusion reactors. This study focuses on the design and optimisation of the VNS Thermal Shield, adopting a multidisciplinary approach to address its thermal and structural behaviours. The Thermal Shield plays a crucial role in protecting superconducting magnets and other cryogenic components by limiting heat transfer from higher-temperature regions of the tokamak to the cryostat, which operates at temperatures between 4 K and 20 K. To ensure both thermal insulation and structural integrity, multiple design iterations were conducted. These iterations aimed to reduce electromagnetic (EM) forces induced during magnet charge and discharge cycles by introducing strategic cuts and reinforcements in the shield design. The optimisation process included the evaluation of various aluminium alloys and composite materials to achieve a balance between rigidity and weight while maintaining structural integrity under EM and mechanical loads. Additionally, an integrated thermal study was performed to ensure effective temperature management, maintaining the shield at an operational temperature of around 80 K. Cooling channels were incorporated to homogenise temperature distribution, improving thermal stability and reducing thermal gradients. This comprehensive approach demonstrates the viability of advanced material solutions and design strategies for thermal and structural optimisation. The findings reinforce the importance of the VNS as a dedicated platform for testing and validating critical fusion technologies under operationally relevant conditions. Full article

For the large-scale fusion magnets of the International Thermonuclear Experimental Reactor (ITER) tokamak, wound with cable-in-conduit conductors, the application of sophisticated numerical models able to analyse the thermal–hydraulic behaviour during plasma scenarios is of paramount importance to guarantee an adequate stability margin during operating conditions. The SuperMagnet code has been developed by CryoSoft with the intent to simultaneously simulate the electrical, thermal and hydraulic phenomena occurring during the operation of superconducting coils. In this work, the SuperMagnet code is applied to analyse the thermal–hydraulic behaviour of the central solenoid of the ITER tokamak under the plasma scenario. The central solenoid (CS) is composed of six modules for a total amount of 240 pancakes. The software is able to tackle the complex structure of the CS and its cryogenic closed loop. In the present work, the circulation pump operation and the heat transfer to the helium bath are investigated. The results presented here show the temperature evolution of the magnet and of the supercritical helium during the plasma scenario, which allows the determination of the operation margin of the CS. Full article

Nuclear fusion reactors are composed of several complex components whose behavior may be not certain a priori. This uncertainty may have a significant impact on the evolution of fault transients in the machine, causing unexpected damage to its components. For this reason, a suitable method for the uncertainty propagation during those transients is required. The Monte Carlo method would be the reference option, but it is, in most of the cases, not applicable due to the large number of required, repeated simulations. In this context, the Polynomial Chaos Expansion has been considered as a valuable alternative. It allows us to create a surrogate model of the original one in terms of orthogonal polynomials. Then, the uncertainty quantification is performed repeatedly, relying on this much simpler and faster model. Using the fast current discharge in the Divertor Tokamak Test Toroidal Field (DTT TF) coils as a reference scenario, the following method has been applied: the uncertainty on the parameters of the Fast Discharge Unit (FDU) varistor disks is propagated to the simulated electrical and electromagnetic relevant effects. Eventually, two worst-case scenarios are analyzed from a thermal–hydraulic point of view with the 4C code, simulating a fast current discharge as a consequence of a coil quench. It has been demonstrated that the uncertainty on the inputs (varistor parameters) strongly propagates, leading to a wide range of possible scenarios in the case of accidental transients. This result underlines the necessity of taking into account and propagating all possible uncertainties in the design of a fusion reactor according to the Best Estimate Plus Uncertainty approach. The uncertainty propagation from input data to electrical, electromagnetic, and thermal hydraulic results, using surrogate models, is the first of its kind in the field of the modeling of superconducting magnets for nuclear fusion applications. Full article

With high real-time performance and a good filtering effect, the wavelet transform-based real-time filter is more commonly used in the fusion magnet power supply. Considering the signal characteristics of the fusion magnet power supply, the sliding time window method is widely used to satisfy the requirements of filtering efficiency. In this paper, an updated half-edge extension method is proposed to improve the boundary effect of sliding window data. Moreover, the hardware-based wavelet loopback structure is proposed to avoid the double computation of long sliding window data, thereby improving real-time performance and satisfying the demands of fusion magnet power supply filtering. The proposed half-edge extension method and the wavelet loopback structure are implemented by designed FPGA and experimented on the Experimental and Advanced Superconducting Tokamak (EAST) fusion magnet power supply system to verify the filtering efficiency. According to the experiment’s results, the proposed half-edge extension method has a better filtering effect, and the proposed wavelet loopback structure decreases the time by 67.4%, which can satisfy the requirements of the power supply system. Full article

The China Fusion Engineering Test Reactor (CFETR), a superconducting magnetic confinement tokamak fusion reactor, will develop a high-performance torus cryopump to pump torus plasma exhaust gas. The inlet valve is one of the key components of the cryopump, and it is used to isolate the cryopump from the plasma for regeneration, to control the pumping speed of the cryopump, and to operate as a pressure relief valve in case of a failure, such as the cryopipe breaking inside the cryopump chamber. This paper presents a novel inlet valve. Ensuring that the design of the inlet valve meets the above requirements will be a challenge. In order to verify the reliability of the inlet valve, its critical components are analyzed and optimized by the Finite Element Method. The effect of the stroke of the inlet valve on pumping performance is then estimated by the Monte Carlo Method, and the pressure profile in the whole flow field is studied to predict the cryopump’s behavior. Finally, the seismic capacity of the optimized inlet valve is analyzed, and the mechanical performance of the inlet valve is shown to meet CFETR design criteria. These design and analysis results will provide technical support and references for the development of the CFETR torus cryopump. Full article

The roadmap to fusion electricity of the European scientific program aims at the realization of the future DEMOnstration (DEMO) fusion power plant. In 2020, the pre-concept design phase of DEMO was completed, defining the concept and characteristics of the main magnets and structures of the machine. Sixteen toroidal D-shaped magnets, six poloidal annular coils and a central solenoid constitute the functioning system core. The reactor is subjected to huge mechanical loads, mainly due to the Lorentz force produced by the combination of the high magnetic fields and operative currents. As a consequence, the loading conditions are extremely demanding for the structural components, and it is crucial to complete a comprehensive static and fatigue assessment before proceeding with the next design iteration. This work focuses on the electromagnetic and structural analyses performed on the toroidal field coil system and its support structures to present the methodological approach developed. Exploiting the finite element method, a three-dimensional model has been defined to obtain the electromagnetic loads on the main time points of the reference plasma scenario and then transfer them to a related 3D structural model, corresponding to the discretization of the electromagnetic one. The structural model was used to obtain the displacement and stress fields at the various time points to perform the mechanical evaluation as well as the fatigue assessment. Full article

The deflection magnet (DM) is the most important component of the Neutral Beam Injection (NBI) system of Experimental Advanced Superconducting Tokamak (EAST), which can magnetically deflect the un-neutralized charged particles after the neutralized process of the beam is extracted from the ion source, and then form a neutral beam injected into the tokamak. Under the operating conditions of the NBI system, by using the thermocouple monitoring system in the experiment, it can be found that the currently operating DM beam collimator has a quite high temperature rise. It is necessary to redesign the DM beam collimator to improve its heat transfer performance. The parallel arrangement of multiple rows of tubes is proposed as the basic method for the redesign of the beam collimator of DM, the thermal-fluid-structure analysis model of this redesign model is established and its temperature field, pressure field and stress field are analyzed. Taking the surface temperature of the beam collimator, the overall dimension after the total tube combination and the pressure drop of the whole structure of collimator as the optimization objectives, and setting the fluid velocity, the tube’s inner diameter and the number of tube rows as the design variables, the optimized design scheme of the beam collimator structure is obtained. From the results of simulation, the new structure can better meet the operation requirements of DM, and its maximum temperature rise is well controlled, which is expected to meet the long pulse operation requirements of the NBI system. The proposed simulation and design optimization method can provide a certain reference for the design and optimization of other high-heat-flux structures in complex large-scale neutral beam systems in the future. Full article

Plasma probes are simple and inexpensive diagnostic tools for fast measurements of relevant plasma parameters. While in earlier times being employed mainly in relatively cold laboratory plasmas, plasma probes are now routinely used even in toroidal magnetic fusion experiments, albeit only in the edge region, i.e., the so-called scrape-off layer (SOL), where temperature and density of the plasma are lower. To further avoid overheating and other damages, in medium-size tokamak (MST) probes are inserted only momentarily by probe manipulators, with usually no more than a 0.1 s per insertion during an average MST discharge of a few seconds. However, in such hot and high-density plasmas, their usage is limited due to the strong particle fluxes onto the probes and their casing which can damage the probes by sputtering and heating and by possible chemical reactions between plasma particles and the probe material. In an attempt to make probes more resilient against these detrimental effects, we tested two graphite probe heads (i.e., probe casings with probes inserted) coated with a layer of electrically isolating ultra-nano-crystalline diamond (UNCD) in the edge plasma region of the Experimental Advanced Superconducting Tokamak (EAST) in Hefei, People’s Republic of China. The probe heads, equipped with various graphite probe pins, were inserted frequently even into the deep SOL up to a distance of 15 mm inside the last closed flux surface (LCFS) in low- and high-confinement regimes (L-mode and H-mode). Here, we concentrate on results most relevant for the ability to protect the graphite probe casings by UNCD against harmful effects from the plasma. We found that the UNCD coating also prevented almost completely the sputtering of graphite from the probe casings and thereby the subsequent risk of re-deposition on the boron nitride isolations between probe pins and probe casings by a layer of conductive graphite. After numerous insertions into the SOL, first signs of detachment of the UNCD layer were noticed. Full article

Superconducting magnets are an invaluable tool for scientific discovery, energy research, and medical diagnosis. To date, virtually all superconducting magnets have been made from two Nb-based low-temperature superconductors (Nb-Ti with a superconducting transition temperature T_c of 9.2 K and Nb₃Sn with a T_c of 18.3 K). The 8.33 T Nb-Ti accelerator dipole magnets of the large hadron collider (LHC) at CERN enabled the discovery of the Higgs Boson and the ongoing search for physics beyond the standard model of high energy physics. The 12 T class Nb₃Sn magnets are key to the International Thermonuclear Experimental Reactor (ITER) Tokamak and to the high-luminosity upgrade of the LHC that aims to increase the luminosity by a factor of 5–10. In this paper, we discuss opportunities with a high-temperature superconducting material Bi-2212 with a T_c of 80–92 K for building more powerful magnets for high energy circular colliders. The development of a superconducting accelerator magnet could not succeed without a parallel development of a high performance conductor. We will review triumphs of developing Bi-2212 round wires into a magnet grade conductor and technologies that enable them. Then, we will discuss the challenges associated with constructing a high-field accelerator magnet using Bi-2212 wires, especially those dipoles of 15–20 T class with a significant value for future physics colliders, potential technology paths forward, and progress made so far with subscale magnet development based on racetrack coils and a canted-cosine-theta magnet design that uniquely addresses the mechanical weaknesses of Bi-2212 cables. Additionally, a roadmap being implemented by the US Magnet Development Program for demonstrating high-field Bi-2212 accelerator dipole technologies is presented. Full article

The qualification and validation of nuclear technologies are daunting tasks for fusion demonstration (DEMO) and power plants. This is particularly true for advanced designs that involve harsh radiation environment with 14 MeV neutrons and high-temperature operating regimes. This paper outlines the unique qualification and validation processes developed in the U.S., offering the only access to the complete fusion environment, focusing on the most prominent U.S. blanket concept (the dual cooled PbLi (DCLL)) along with testing new generations of structural and functional materials in dedicated test modules. The venue for such activities is the proposed Fusion Nuclear Science Facility (FNSF), which is viewed as an essential element of the U.S. fusion roadmap. A staged blanket testing strategy has been developed to test and enhance the DCLL blanket performance during each phase of FNSF D-T operation. A materials testing module (MTM) is critically important to include in the FNSF as well to test a broad range of specimens of future, more advanced generations of materials in a relevant fusion environment. The most important attributes for MTM are the relevant He/dpa ratio (10–15) and the much larger specimen volumes compared to the 10–500 mL range available in the International Fusion Materials Irradiation Facility (IFMIF) and European DEMO-Oriented Neutron Source (DONES). Full article