Search Results (191)

The Multi-Mode Model (MMM) is a physics-based anomalous transport model integrated into TRANSP for predicting electron and ion thermal transport, electron and impurity particle transport, and toroidal and poloidal momentum transport. While MMM provides valuable predictive capabilities, its computational cost, although manageable for standard simulations, is too high for real-time control applications. MMMnet, a neural network-based surrogate model, is developed to address this challenge by significantly reducing computation time while maintaining high accuracy. Trained on TRANSP simulations of DIII-D discharges, MMMnet incorporates an updated version of MMM (9.0.10) with enhanced physics, including isotopic effects, plasma shaping via effective magnetic shear, unified correlation lengths for ion-scale modes, and a new physics-based model for the electromagnetic electron temperature gradient mode. A key advancement is MMMnet’s ability to predict all six transport coefficients, providing a comprehensive representation of plasma transport dynamics. MMMnet achieves a two-order-of-magnitude speed improvement while maintaining strong correlation with MMM diffusivities, making it well-suited for real-time tokamak control and scenario optimization. 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

Copper alloys are critical heat sink materials for fusion reactor divertors due to their high thermal conductivity (TC) and strength, yet their performance under extreme particle bombardment and heat fluxes in future tokamaks requires enhancement. While neutron-induced transmutation helium affects the properties of copper, the atomistic mechanisms linking helium bubble size to thermal transport remain unclear. This study employs non-equilibrium molecular dynamics (NEMD) simulations to isolate the effect of bubble diameter (10, 20, 30, 40 Å) on TC in copper, maintaining a constant He-to-vacancy ratio of 2.5. Results demonstrate that larger bubbles significantly impair TC. This reduction correlates with increased Kapitza thermal resistance and pronounced lattice distortion from outward helium diffusion, intensifying phonon scattering. Phonon density of states (PDOS) analysis reveals diminished low-frequency peaks and an elevated high-frequency peak for bubbles >30 Å, confirming phonon confinement and localized vibrational modes. The PDOS overlap factor decreases with bubble size, directly linking microstructural evolution to thermal resistance. These findings elucidate the size-dependent mechanisms of helium bubble impacts on thermal transport in copper divertor materials. Full article

Negative Ion Source Neutral beam Injection (NNBI), as a critical auxiliary heating system for magnetic confinement fusion devices, directly affects the plasma heating efficiency of tokamak devices through the reliability of its beam source system. The single-shot experiment constitutes a significant experimental program for NNBI. This study addresses the frequent equipment failures encountered by the NNBI beam source system during a cycle of experiments, employing fault tree analysis (FTA) to conduct a systematic reliability assessment. Utilizing the AutoFTA 3.9 software platform, a fault tree model of the beam source system was established. Minimal cut set analysis was performed to identify the system’s weak points. The research employed AutoFTA 3.9 for both qualitative analysis and quantitative calculations, obtaining the failure probabilities of critical components. Furthermore, the F-V importance measure and mean time between failures (MTBF) were applied to analyze the system. This provides a theoretical basis and practical engineering guidance for enhancing the operational reliability of the NNBI system. The evaluation methodology developed in this study can be extended and applied to the reliability analysis of other high-power particle acceleration systems. Full article

In the decommissioning of nuclear facilities, activated steel often contains radionuclides such as ⁵⁵Fe and ⁶³Ni, which are categorized as hard-to-measure due to their emission of only low-energy beta particles or X-rays. In samples exhibiting very low radioactivity, close to background levels, a large quantity of steel must undergo extensive physical and chemical processing to achieve the Minimum Detectable Activity Concentration (MDC) necessary for clearance, recycling, or reuse. Italian regulations set particularly stringent clearance levels for these radionuclides (1 Bq/g for both ⁵⁵Fe and ⁶³Ni), significantly lower than those specified in the EU Directive 2013/59 (1000 Bq/g for ⁵⁵Fe and 100 Bq/g for ⁶³Ni). Additionally, Italian authorities may enforce even stricter limits depending on specific circumstances. The analytical challenge is compounded by the presence of large amounts of non-radioactive Fe and Ni, which can cause color quenching, further extending analysis times. This study presents a reliable and optimized method for the quantitative determination of ⁵⁵Fe and ⁶³Ni in steel samples with activity levels approaching regulatory thresholds. The methodology was specifically developed and applied to steel from the Frascati Tokamak Upgrade (FTU) facility, under decommissioning by ENEA. The optimization process demonstrated that achieving the required MDCs necessitates acquisition times of approximately 5 days for ⁵⁵Fe and 6 h for ⁶³Ni, ensuring compliance with stringent regulatory requirements and supporting efficient laboratory workflows. Full article

The mechanisms by which rotation influences zonal flows (ZFs) in plasma are incompletely understood, presenting a significant challenge in the study of plasma dynamics. This research addresses this gap by investigating the role of non-inertial effects—specifically centrifugal and Coriolis forces—on Geodesic Acoustic Modes (GAMs) and ZFs in rotating tokamak plasmas. While previous studies have linked centrifugal convection to plasma toroidal rotation, they often overlook the Coriolis effects or inconsistently incorporate non-inertial terms into magneto-hydrodynamic (MHD) equations. In this work, we derive self-consistent drift-ordered two-fluid equations from the collisional Vlasov equation in a non-inertial frame, and we modify the Hermes cold ion code to simulate the impact of rotation on GAMs and ZFs. Our simulations reveal that toroidal rotation enhances ZF amplitude and GAM frequency, with Coriolis convection playing a critical role in GAM propagation and the global structure of ZFs. Analysis of simulation outcomes indicates that centrifugal drift drives parallel velocity growth, while Coriolis drift facilitates radial propagation of GAMs. This work may provide valuable insights into momentum transport and flow shear dynamics in tokamaks, with implications for turbulence suppression and confinement optimization. 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

In axisymmetric fusion devices like tokamaks, the winding of the magnetic field is characterized by its safety profile $q=q_{\mathbf{B}}$. Similarly, the winding of the current density field is characterized by $q_{\mathbf{J}}$. Currently, the relationship between $q_{\mathbf{B}}$ and $q_{\mathbf{J}}$ profiles and their effect on tokamak plasma confinement properties remains unexplored, as the $q_{\mathbf{J}}$ profile is neither computed nor considered. This study presents a reconstruction of the current density winding profile from experimental data in the quiescent H-mode. The topology analysis derived from $(q_{\mathbf{B}},q_{\mathbf{J}})$ was carried out using Hamada coordinates. It shows a large central plasma region unaffected by current filamentation-driven resonant magnetic perturbations, while the outer region harbors a spectrum of magnetic resonant modes, induced by current filaments located within the core plasma, which degrade peripheral confinement. These results suggest a QH-mode signature pattern needing further validation with additional data. Implementing $(q_{\mathbf{B}},q_{\mathbf{J}})$ real-time monitoring could provide insights into tokamak confinement regimes with significant implications. Full article

The present study focuses on the development of a diagnostic system for measuring radiated power and core soft X-ray intensity emissions with the goal of detecting a broad spectrum of photon energies emitted from the central plasma region of the DEMO tokamak. The principal objective of the diagnostic apparatus is to deliver a comprehensive characterization of the radiation emitted by the plasma, with a particular focus on estimating the radiated power from the core region. This measurement is essential for determining and monitoring the power crossing the separatrix, which is a critical parameter controlling overall plasma performance. Since diagnostics rely on line-integrated measurements, the application of tomographic reconstruction techniques is necessary to extract spatially resolved information on core plasma radiation. This contribution presents the development of numerical algorithms addressing the problem of radiation tomography reconstruction. A robust and computationally efficient method is proposed for reconstructing the spatial distribution of plasma radiated power, with a view toward enabling real-time applications. The reconstruction methodology is based on a linear model formulated using a set of predefined basis functions, which define the radiation distribution within a specified plasma cross-section. In the initial stages of emissivity reconstruction in tokamak plasmas, it is typically assumed that the radiation distribution is dependent on magnetic flux surfaces. As a baseline approach, the plasma radiative properties are considered invariant along these surfaces and can thus be represented as one-dimensional profiles parameterized by the poloidal magnetic flux. Within this framework, the reconstruction method employs an approximation model utilizing three sets of basis functions: (i) polynomial splines, as well as Gaussian functions with (ii) sigma parameters and (iii) position parameters. The performance of the proposed method was evaluated using two synthetic radiated power emission phantoms, developed for the DEMO plasma scenario. The results indicate that the method is effective under the specified conditions. Full article

The exploitation of fusion energy in tokamak reactors relies on efficient and reliable tritium management. The tritium needed to sustain the deuterium–tritium fusion reaction is produced in the Li-based blanket surrounding the plasma chamber, and, therefore, the effective extraction and purification of the tritium bred in the Li-blankets is needed to guarantee the tritium self-sufficiency of future fusion plants. This work introduces a new technology for the extraction of tritium from the Pb–Li eutectic alloy used in liquid blankets. Process units based on the concept of Membrane Gas–Liquid Contactor (MGLC) have been studied for the extraction of tritium from the Pb–Li in the Water Cooled Lithium Lead blankets of the DEMO reactor. MGLC units have been preliminarily designed and then compared in terms of the permeation areas and sizes with the tritium extraction technologies presently under study, namely the Permeator Against Vacuum (PAV) and the Gas–Liquid Contactors (GLCs). The results of this study show that the DEMO WCLL tritium extraction systems using MGLC require smaller permeation areas and quicker permeation kinetics than those based on PAV (Permeator Against Vacuum) devices. Accordingly, the MGLC extraction unit exhibits volumes smaller than those of both PAV and GLC. Full article

The future fusion power plant EU DEMO will generate its own tritium fuel through the use of segmented breeding blankets (BBs), which must be replaced from time to time due to material damage caused by high-energy neutrons from the plasma. A vertical maintenance architecture has been proposed, using a robotic remote handling tool (transporter) to disengage the 180 $\mathrm{t}$ and 125 $\mathrm{t}$ outboard and inboard segments and manipulate them through an upper port. Safe disengagement without damaging the support structures requires the use of high-capacity tilting joints in the transporter. The trolley tilting mechanism (TTM) is proposed as a novel, compact, high-capacity robotic joint consisting of a five-bar spatial mechanism integrated in the BB transporter trolley link. A kinematic model of the TTM is established, and the analytical input–output relationships, including the position-dependent transmission ratio, are derived and used to guide the design and optimization of the mechanism. The model predictions are compared to an ADAMS multibody simulation and to the results of an experiment conducted on a down-scaled prototype, both of which validate the model accuracy. Full article

The paper presents a conceptual study of a neutron source based on a spherical tokamak (ST). The plasma scenario chosen for the ST is non-thermal fusion (hot ion mode), which is extensively used on machines like JET and TFTR deuterium–tritium (DT) experiments, which seems suited for low fusion gain reactors. As demonstrated in experiments, this scenario is a robust tool for neutron production. Starting from a new scaling law of energy confinement tested, approximately, on ST40 spherical tokamak, the parameters of a 15 MW ST DT fusion reactor (ST180) are derived, and a preliminary radial build of the machine is established. 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

The presence of a bright spot structure in the divertor region during the discharge process, indicative of localized overheating, has been observed through multi-band and high-speed endoscope diagnostic on the Experimental Advanced Superconducting Tokamak (EAST). This localized deposition of hyperthermal heat flux can lead to erosion and melting of the target plate material, thereby posing a significant risk to the safe operation of the device. Moreover, it may introduce impurities into the main plasma, negatively impacting plasma performance. Therefore, real-time monitoring of the divertor and rapid identification of localized overheating regions during experiments are crucial. In this context, this paper proposes an improved DeepLabv3+-based highlight structure image-segmentation algorithm, which uses minimum value, image difference method, and Prewitt operator for dataset preprocessing. In order to realize the rapid identification of local overheated regions, this paper introduces the application of the improved DeepLabv3+ neural network algorithm based on MobileNetV2 as the backbone network in the bright spot structure segmentation task for the first time. The results show that the algorithm achieves a 65.36% average crosslinking rate (mIoU), 78.75% accuracy, 0.78 s per-iteration processing time, and 22.4 MB parameter size. This provides substantial advantages in terms of reduced computing and memory resources and real-time detection performance. Ultimately, the method proposed in this paper enables the rapid identification of the bright spot structure in the localized overheating region of the divertor on the EAST; it identifies areas of overheating and prevents damage to the divertor or other critical components due to overheating, ensuring safe operation of the device. Full article

Nuclear fusion is a promising energy source. The International Thermonuclear Experimental Reactor aims to study the feasibility of tokamak-type reactors and test technologies and materials for commercial use. One major challenge is developing materials for the reactor’s divertor, which supports high thermal flux. Tungsten was chosen as the plasma-facing material, while a CuCrZr alloy will be used in the cooling pipes. However, the gradient between the working temperatures of these materials requires the use of a thermal barrier interlayer between them. To this end, refractory high-entropy (CrFeTiTa)₇₀W₃₀ and VFeTiTaW alloys were prepared by mechanical alloying and sintering, and their thermal and irradiation resistance was evaluated. Both alloys showed phase growth after annealing at 1100 °C for 8 days, being more pronounced for higher temperatures (1300 °C and 1500 °C). The VFeTiTaW alloy presented greater phase growth, suggesting lower microstructural stability, however, no new phases were formed. Both (as-sintered) alloys were irradiated with Ar⁺ (150 keV) with a fluence of 2.4 × 10²⁰ at/m², as well as He⁺ (10 keV) and D⁺ (5 keV) both with a fluence of 5 × 10²¹ at/m². The morphology of the surface of both samples was analyzed before and after irradiation showing no severe morphologic changes, indicating high irradiation resistance. Additionally, the VFeTiTaW alloy presented a lower deuterium retention (8.58%) when compared to (CrFeTiTa)₇₀W₃₀ alloy (14.41%). Full article