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Any means to control gravity like electromagnetism is currently out of reach by many orders of magnitude even under extreme laboratory conditions. Some often poorly executed experiments or pseudoscience theories appear from time to time claiming for example anomalous forces from capacitors that suggest a connection between the two fields. We developed novel and high resolution horizontal-, vertical- and rotation-balances that allow to test electric devices completely shielded and remotely controlled under high vacuum conditions to perform the first in-depth search for such a coupling using steady fields. Our testing included a variety of capacitors of different shapes and compositions as well as for the first-time solenoids and tunneling currents from Zener diodes and varistors. A comprehensive coupling-scheme table was used to test almost all combinations including capacitors and solenoids with permittivity and permeability gradients as well as capacitors and varistors within crossed magnetic fields. We also tested a crossed-coil producing helical magnetic field lines as well as interactions between a pair of shielded toroidal coils to look for proposed extensions to Maxwell's equations. No anomalous forces or torques down to the nano-Newton or nano-Newton-Meter range were found providing new limits many orders of magnitude below previous assessments ruling out claims or theories and providing a basis for future research on the topic. [ABSTRACT FROM AUTHOR]

We propose a scalable system of compact, superconducting neutron monitors, which can be embedded in any existing cryogenic infrastructure of a fusion system. The pixel-based nature of the detectors allows them to be placed at intervals following the circumference of a cooled zone, e.g., a field coil, thus allowing for a tomographic measurement of the neutron flux surrounding the plasma. An early stage prototype of the superconducting bolometer is described, and the key results of a previous feasibility study of this prototype performed with cold neutrons are summarized. The bolometer can be adapted for use with fast neutrons by altering the composition and geometry of the neutron-to-heat conversion layer. This paper describes the initial feasibility considerations for implementation in a superconducting tokamak. The sensor is based on a high-temperature superconductor, making it possible to select the operation temperature in the range 1–90 K. Neutron flux numbers were found using the ITER MCNP reference model, and these were embedded in a TOPAS model to find the expected signal measured by the bolometer at the position of a toroidal field coil. The results at the coil position indicate suitable operation levels in terms of the magnitude of the measured signal, with a measurable signal of several ohm, which is much smaller than the saturation energy of the detector. Radiation hardness is estimated and found to be on the order of at least 40 years for the relevant radiation levels. The upcoming investigation activities of the project are described for both radiation testing and analytical modeling. [ABSTRACT FROM AUTHOR]

This paper presents a study on the magnetic and thermal behaviors of a planar toroidal transformer, comprising two planar toroidal coils. In our configuration, the primary coil consists of twenty turns, while the secondary coil consists of ten turns. This design combines the advantages of both toroidal and planar transformers: it employs flat coils, akin to those utilized in planar transformers, while retaining a toroidal shape for its magnetic core. This combination enables leveraging the distinctive characteristics of both transformer types. This study delves into electromagnetic and thermal behaviors. Electromagnetic behavior is elucidated through Maxwell's equations, offering insights into the distribution of magnetic fields, potentials, and electric current densities. Fluid flow is modeled via the Navier–Stokes equations. By coupling these equation sets, a more comprehensive and accurate portrayal of the thermal phenomena surrounding electrical equipment is attained. Such research is invaluable in the design and optimization of electrical systems, empowering engineers to forecast and manage thermal effects more efficiently. Consequently, this aids in enhancing the reliability, durability, and performance optimization of electrical equipment. The mathematical model was solved using the finite element method integrated into the COMSOL Multiphysics software v. 6.0. The COMSOL Multiphysics simulation showed correct behavior of potential, electric field, current density, and uniformly distributed temperature. In addition, this planar toroidal coil transformer model offers many advantages, such as small dimensions, high resonance frequency, and high operating reliability. This study made it possible to identify the range of its optimal functioning. [ABSTRACT FROM AUTHOR]

The future fusion reactor devices built by Institute of Plasma Physics, Chinese Academy of Science will be a compact burning plasma facility to fill the gap between ITER and the China Fusion Engineering Test Reactor, with the mission of studying the deuterium–tritium plasma in steady-state operation. The winding-package (WP) of its toroidal field coil is graded into high-field WP and low-field WP, and the high critical current density (J c) Nb3Sn strand will be applied to the high-field WP based on the current design. The thermo-magnetic instability is the main issue in the application of high- J c Nb3Sn strand, which may induce premature quenching in the low field regions. This issue can be improved by reducing the effective filament diameter (d eff) to suppress flux jumps and increasing the residual resistivity ratio (RRR) to enhance the release of heat generated by flux jumps. However, in the conductor manufacturing process, local plastic deformations (indentation) of strands can impact the sub-element layout and degrade the RRR of the strand, which would increase again the risk on instability. In this study, magnetization measurements and V – I tests were performed on samples with different indentation depths. The hysteresis loss and d eff of the indented sample was obtained from the test results. The impact of local plastic deformation on the thermo-magnetic instability of two types of Nb3Sn strand was confirmed by cross-sectional metallographic observation and quantitative microstructure analysis. It was shown that flux jumps were suppressed at indentation depths below 0.3 mm. Further indentation increase leads to severe flux jumps. [ABSTRACT FROM AUTHOR]

The recent Federici et al (2024 Nucl. Fusion 64 036025) article makes the argument that higher magnetic fields cannot reduce the size and cost of a tokamak-based fusion power plant due to: structural considerations of the toroidal field (TF) coils, the required thickness for neutron shielding and the blanket, and challenges with heat exhaust in the divertor. This conclusion is based on a series of assumptions that are design decisions made by the authors, not fundamental limits on physics or engineering. This Comment demonstrates that the conclusions of Federici et al are invalid if one makes different design choices and that its results are therefore not broadly generalizable. [ABSTRACT FROM AUTHOR]

This paper presents the verification of results of the stress analysis of the Globus-M2 spherical tokamak magnet system on the basis of measurements carried out in experiments. The magnet system is composed of toroidal and poloidal field coils and intercoil and supporting structures and is subjected to the impact of electromagnetic (EM) loads during an experiment. In the series of experiments with close to maximum values of the plasma current and toroidal field, the toroidal field coil displacements in the toroidal direction under EM loads with respect to its natural position were measured using a high-speed camera. The EM loads were determined using a 3D model of the Globus-M2 tokamak magnet system from the currents of the toroidal and poloidal field coils and plasma measured in experiments. These loads were used for the stress analysis of the Globus-M2 magnet system in the ANSYS software. Results of this work show that the approach consisting of two stages—the reconstruction of the EM loads and the determination of the mechanical response of the structure—was successfully verified for the EM system of the Globus-M2 and can be used for other tokamaks. [ABSTRACT FROM AUTHOR]

In power electronic systems, electromagnetic radiation of the near field is detrimental because it generates electromagnetic interference in static converters ultimately leading to the degradation of the EMC filter performance. This paper presents an experimental investigation into magnetic radiation generated by magnetic components, specifically toroidal inductors, in the time domain (TD) using an oscilloscope and in the frequency domain (FD) via a spectrum analyzer. The method employed to quantify the magnetic field is the near field scan (NFS) using magnetic probes. This approach enables us to identify the areas with the highest magnetic field emissions from the toroidal inductance across its three components (HX, HY, HZ). Initially, our investigation focused on two key parameters of the pulse signal. First, we examined the impact of varying the pulse signal's frequency while keeping the voltage constant. Next, we analyzed the effect of variations in the amplitude of the pulse signal while keeping a constant frequency. Following these experiments, we generated magnetic field maps in the frequency domain (FD) above the surface of the inductor, specifically focusing on the HZ component. The scanning step is equal to 0,5 cm. The scanning area is equal to (x, y) = (4 cm, 4 cm). [ABSTRACT FROM AUTHOR]

This article describes the design and configuration of the DEMO Breeding Blanket (BB) feeding pipes inside the upper port. As large BB segments require periodic replacement via the upper vertical ports, the space inside the upper port needs to be maximized. At the same time, the size of the upper port is constrained by the available space in between the toroidal field coils and the required space to integrate a thermal shield between the vacuum vessel (VV) port and the coils. The BB feeding pipes inside the vertical port need to be removed prior to BB maintenance, as they obstruct the removal kinematics. Since they are connected to the BB segments on the top and far from their vertical support on the bottom, the pipes need to be sufficiently flexible to allow for the thermal expansion of the BB segments and the pipes themselves. The optimization and verification of these BB pipes inside the upper port design are critical aspects in the development of DEMO. This article presents the chosen pipe configuration for both BB concepts considered for DEMO (helium- and water-cooled) and their structural verification for some of the most relevant thermal conditions. A 3D model of the pipes forest, both for the Helium-Cooled Pebble Bed (HCPB) and Water-Cooled Lithium Lead (WCLL) concepts, has been developed and integrated inside the DEMO Upper Port (UP), Upper Port Ring Channel, and Upper Port Annex (UPA). A preliminary structural analysis of the pipeline was carried out to check the structural integrity of the pipes, their flexibility against the thermal load, their internal pressure, and the deflection induced by the thermal expansion of the BB segments. The results showed that the secondary stress on the hot leg of the HCPB pipeline was above the limit, suggesting future improvements in its shape to increase the flexibility. Moreover, the WCLL concept did not have a critical point in terms of the secondary stress on the pipeline, since the thicknesses and the diameters of these pipes were smaller than the HCPB ones. [ABSTRACT FROM AUTHOR]

Copyright of Electric Machines & Control / Dianji Yu Kongzhi Xuebao is the property of Electric Machines & Control and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

The structure, the phase composition, and the dynamics of phase transformations of nanosystems based on zirconium dioxide ZrO2(CaO) synthesized by the plasma-chemical method are studied. In order to lower the sintering temperatures and improve the physical and mechanical properties of the ceramics, the ZrO2–CaO nanocrystalline system is subjected to activation sintering in an external magnetic field. A constant magnetic field inside the furnace volume is created by a toroidal coil with a special symmetrized winding. The ceramics sintered in a constant magnetic field of B = 0.02–1 T with a given C3 symmetry at 1200–1400°C exhibits improved physical and mechanical characteristics and more perfect crystalline microstructures. [ABSTRACT FROM AUTHOR]

The GLOBSYS code was developed for analysis and prediction of parameters of the Globus-M2 tokamak and its modifications. In [1], preliminary selection of correlations which connect physical and technical parameters was made. In this paper, the verification of the code using the achieved and predicted data from the installations NSTX, NSTX-U, MAST, MAST-U, and ST40 is given. As a whole, there is good agreement between simulations and plasma parameters at the discharge plateau. The best agreement is observed if ITER confinement scaling is used for energy confinement time with the enhancement factor Hy, 2 = 1–1.2. Simulations with other confinement scalings (Globus-2021, NSTX scalings) give good agreement with plasma parameters for the toroidal field Bt0 ~ 0.5 T. For increasing Bt0, more optimistic predicted plasma parameters are obtained for the Globus-2021 and NSTX scalings in comparison with the ITER confinement scaling. The condition of reaching the plasma quasistationary regime (or the time of establishment of quasistationary plasma profiles τL/R) is estimated for NSTX, NSTX-U, MAST, MAST-U and ST40 discharges. This time is compared with two technical restrictions, which are connected with the times of toroidal field coil heating and poloidal flux capacity. Verification of the GLOBSYS code using the data from the aforementioned spherical tokamaks is the basis for the prediction of parameters of the next step of Globus-M program. [ABSTRACT FROM AUTHOR]

The T-15MD tokamak with the warm toroidal field coils, elongated plasma cross section, and a low aspect ratio was commissioned at the National Research Center Kurchatov Institute in 2021. An economically viable fusion reactor requires long discharges or even a steady-state mode. The short pulse of the toroidal field is among the main obstacles on the way to the steady state. Therefore, as a next step after T-15MD, it is proposed at the NRC Kurchatov Institute to consider the design of a superconducting tokamak (SCT), which holds the basic geometric parameters of the T-15MD: R/a = 1.5 m/0.67 m with a field on the axis of B0 < 5 T and a long pulse of current of Ip < 5 MA. The focus of the paper is the system of the toroidal magnetic field. A three-layer design of the toroidal field coil is proposed: an inner layer of a high-temperature superconductor (HTSC), a Nb3Sn middle layer, and a NbTi outer layer. On the basis of structural strength calculations, candidate materials for coil cases were selected. For cooling the coils, a semi-longitudinal pumping of low-pressure liquid helium is proposed. The design of the HTSC cable in conduit consisting of two halves is considered in detail. A mesh cryostat design that provides a convenient access for diagnostics and plasma heating is proposed. SCT systems similar to those used in the T-15MD are analyzed. [ABSTRACT FROM AUTHOR]

A principal part of the Comprehensive Research Facility for Fusion Technology (CRAFT) is the toroidal field (TF) coil project, which aims at fabricating a prototype TF coil for China Fusion Engineering Test Reactor (CFETR). The TF coil is composed of 3-D-shaped subwinding packs (sub-WPs) enclosed in a stainless coil case, and the conductor core of medium and high field sub-WPs is Nb3Sn, following the “Winding & Reaction” manufacturing process. After heat treatment of the conductor, turn insulation will be automatically wrapped. Strictly restricted by the brittleness of the heat-treated Nb3Sn, the strain must be controlled during the wrapping process. Presented in this article is a description of the CRAFT TF coil and its turn insulation structure, followed by descriptions of the structural design of the wrapping unit and mathematical models about wrapping control of the wrapping machine. Turn insulation wrapping trials are performed with one-fourth part of the wrapping unit, and tapes with a fixed width and a uniform steel tube are used as specimens, verifying the correctness of the structure and models. The one-fourth part of the wrapping unit has been fabricated in March 2022 and it is foreseen that the fabrication will be finished in June 2022. [ABSTRACT FROM AUTHOR]

Permanent magnets provide a possible solution to simplify complicated stellarator coils. A prototype permanent magnet stellarator structure, PM4Stell, has been funded to demonstrate the technical feasibility of using permanent magnets to create the shaping field of a stellarator. Permanent magnets in uniform cubes with three polarization directions will be carefully placed to generate the required magnetic field for a National Compact Stellarator eXperiment-like equilibrium together with planar toroidal field coils. Discrete magnets will be glued together and inserted into a "post-office-box-like" supporting structure. Electromagnetic and structural analyses have been done to validate the design. Error field correction magnets will be used to shim possible error fields. The design efforts of the prototype permanent magnet stellarator structure are discussed. [ABSTRACT FROM AUTHOR]

In this work, we propose a cost-effective procedure to obtain magnetic single cotton yarns by employing commercially available magnetic powders and adhesives. Two different indirect magnetization methods are used on the yarns coated with several mixtures containing barium hexaferrite (BaFe12O19) as hard magnetic powder. The BH mass percentage is varied between 10% and 50%. Specific textile coating binders such as polyurethane and polyvinyl acetate are used as adhesives. A constant amount of 5% glycerol is employed as plasticizer agent. An in-house developed laboratory equipment is used to produce the magnetic cotton yarns. The equipment allows primary orientation of the magnetic particles along external field lines and subsequent magnetization until saturation. Our studies show that the diameter of the coated yarns is directly dependent on the amount of magnetic powder in the coating solution. Thus, the magnetic yarn diameter increases by 30% when compared to the 356 µm diameter of the reference (uncoated) yarn for a barium hexaferrite mass percentage of 50%. Also, our studies reveal that increasing the magnetic properties (residual magnetism and coercive field intensity) of the composite yarns is possible only by increasing the mass percentage of the magnetic powder content. The highest values of the magnetic properties have been measured when neodymium permanent magnets (NdFeB) were used for the magnetization instead of a toroidal coil. The residual magnetization and saturation increase with the amount of barium hexaferrite embedded in the textile yarn. However, the increase in mass percentage is limited by the degradation of the yarn properties which allow them to be used for textile applications. SEM images of the coated yarns reveal a relatively uniform deposition of magnetic layer on the reference cotton yarn. [ABSTRACT FROM AUTHOR]

• Context within EUROfusion: the work is conducted within the EUROfusion framework, emphasizing its significance within the fusion research community. • Addressing Fundamental Issues: it aims to tackle fundamental issues crucial for achieving energy extraction from fusion reactors, indicating its importance in advancing fusion technology. • Plasma Aspect Ratio and PF system size: choosing a significantly high aspect ratio for the plasma results in a configuration where the PF coil system is too distant and subsequently too large to control the plasma in a conventional manner. • Machine Assembly Configuration: it proposes an alternative machine assembly configuration, highlighting the importance of proximity between poloidal magnets and the plasma. The PF system is now placed inside the TF system. Both systems are based on HTS magnets. • Feasibility Analysis: the paper includes a feasibility analysis of the proposed solution, providing details on a proposed overall fabrication and assembly sequence for the tokamak. The paper introduces a new approach to assemble tokamak machines applied to the plasma-based Volumetric Neutron Source (VNS) device, addressing challenges related to machine topology. The VNS, designed as a compact tokamak, serves as a qualification testbed for in-vessel fusion components, especially the breeding blanket. The project aims to bridge the technology gap between ITER and DEMO. The machine size has been minimized to maximize the Neutron Wall Load (NWL). This results in a small, dense plasma and, due to the massive neutron shielding structure, a comparably large distance between the plasma and the equilibrium magnets, challenging the primary stabilization and shaping of the plasma. To overcome this, a magnet system architecture is proposed for VNS in which the poloidal field (PF) coil system is placed inside the TF coils. An assembly procedure is proposed, where the superconducting PF coils are first completed and positioned around the vacuum vessel, followed by the in-situ winding and assembly of the toroidal field coil system. The main magnetic cage components of the machine are based on High-Temperature Superconductors (HTS) to avoid the need for heat treatment. The paper presents the advantages of the magnet system architecture, and the preliminary feasibility study of the machine assembly process. [ABSTRACT FROM AUTHOR]

• Sizing of major DEMO tokamak components. • Identification of relation between aspect ratio and magnetic field. • Identification of most compact DEMO design point. • Discards the choice of DEMO as a spherical tokamak. • Discards the choice of DEMO with a high aspect ratio and a high magnetic field. The design point that had been chosen for EU DEMO in 2016 is reviewed here and a modification is proposed with a lower aspect ratio. Previously the same aspect ratio, A, was chosen for EU DEMO as in major tokamak experiments including ITER (A = 3.1), and, to rely on mature technology, a peak magnetic field no greater than 13 T was considered. Here we do not consider these limitations recognizing the recent commissioning of JT60-SA with A = 2.5 and the successful recent operation of a model coil at a field of >20 T. EU DEMO must have a burning plasma and meet performance requirements relevant to a fusion power plant - at present, 2 GW fusion power and 2 h pulse length. The better plasma confinement at higher magnetic field allows reaching this condition in a smaller plasma. Thus, increasing the magnetic field appears as an obvious strategy to reduce the machine size. We confirmed though previous observations that the choice of a high magnetic field is associated with a large aspect ratio, mainly to generate space for the larger TF coils. In practice the magnetic field strength on DEMO-size TF coils is limited to ∼12 T by the high electromagnetic loads. Also, the extreme heat flux on the divertor increases further with the magnetic field. Hence the magnetic field on the plasma axis is limited in EU DEMO to ∼5.4 T, its aspect ratio to approximately 3. The limiting factor to lowering the aspect ratio is the space on the inboard side. This is primarily driven by the requirement to integrate the central solenoid to drive the plasma current inductively. Our literature review suggests that non-inductive plasma scenarios, as considered in most power plant studies in literature, are optimistic and not sufficiently supported by experimental results. Also, the space required for the superconducting toroidal field coils, the tritium breeding blanket, and the neutron shield is substantial. For a DEMO device the space on the inboard side becomes insufficient for aspect ratios below ∼2.6. We therefore conclude the aspect ratio of EU DEMO should be chosen within the range ∼2.6 - ∼3.0 trading-off lower magnetic field and lower divertor heat loads against machine compactness. [ABSTRACT FROM AUTHOR]

In a typical Nb3Sn cable for fusion applications, the knowledge of the strain state of Nb3Sn filaments is important to predict and interpret the conductor performance in operation. Amongst the relevant strain sources, the predominant is the thermal strain that arises during the cool-down of the conductor to operating temperature, due to the different expansion of Nb3Sn and copper in the cable and the stainless steel conduit. This work addresses the characterization of Nb3Sn thermal strain in the React & Wind prototype (RW2) developed by EPFL-SPC for the toroidal field coil and the central solenoid of the EUROfusion DEMO fusion reactor. AC susceptibility measurements were performed on the RW2 prototype to assess the thermal strain distribution in the cable cross-section, before and after undergoing electromagnetic and thermal cyclic loading. The results of the analysis have shown that RW2 exhibits a significantly lower absolute mean strain value (about −0.3%), compared to many Wind & React ITER conductor samples investigated with the same method. In addition, the analysis highlighted that the assessed thermal strain distribution does not vary significantly before and after the electromagnetic and thermal cycles, confirming the small performance degradation in the RW2 conductor that has been recently reported. [ABSTRACT FROM AUTHOR]

Both linear and quasi-linear aspects of the plasma response to the resonant magnetic perturbation (RMP) field are numerically investigated for various H-mode scenarios in ITER, covering the pre-fusion power operation and the fusion power operation phases. Linear response computations for eight ITER scenarios, with varying plasma current and toroidal magnetic field, reveal that the best coil current phasing for controlling the type-I edge localized modes (ELMs) scales roughly linearly with the edge safety factor. The coil phasing is defined as the relative toroidal phase of the coil currents between different rows, for a given toroidal harmonic. Quasi-linear initial value simulation, which is the focus of the present study, shows that application of the n = 3 (n is the toroidal mode number) RMP field has a minimum side effect on the plasma core momentum confinement but potentially a large effect on the global particle transport. Generally, the RMP field with the best (worst) coil phasing for ELM control produces the strongest (weakest) effect on the plasma edge flow and the overall density. This robustly holds for all eight ITER scenarios. Consequently, in order to minimize the RMP induced side effects while achieving ELM control (suppression) in ITER, a compromise is necessary in choosing the coil current configuration. [ABSTRACT FROM AUTHOR]

The helium-cooled pebble bed (HCPB) breeding blanket (BB) is one of the two driver-blanket candidates for the European DEMO fusion reactor. Recent design activities were focused, among other objectives, on the achievement of an efficient shielding system to adequately protect the vacuum vessel (VV) and toroidal field coils (TFCs). Several shielding options have been studied in terms of architecture (e.g., in-BB shield and ex-BB shield) and materials (e.g., B4C, WC, WB, YHx, and ZrHx). In this study, the B4C material was selected as the most attractive option considering not only shielding performance but also availability, industrialization, experience, and cost factors. Subsequently, we performed a parametric study by implementing different thicknesses of a B4C external shield and reporting information of its effect on shielding performance, structural behavior, swelling and tritium breeding. Furthermore, a detailed structure for the VV was developed considering an internal layered configuration comprising steels/water with different boron contents. Corresponding shielding analyses were conducted regarding influence on neutron attenuation when implementing such a VV structure for both the baseline consolidated design of the HCPB and one of the previously developed and improved BSS configurations. The most critical responses (neutron flux and dpa) were fully established only using 10 cm B4C and an improved VV configuration. [ABSTRACT FROM AUTHOR]

As the conventional voltage and current (VI) probes widely used in plasma diagnostics have separate voltage and current sensors, crosstalk between the sensors leads to degradation of measurement linearity, which is related to practical accuracy. Here, we propose a VI probe with a floating toroidal coil that plays both roles of a voltage and current sensor and is thus free from crosstalk. The operation principle and optimization conditions of the VI probe are demonstrated and established via three-dimensional electromagnetic wave simulation. Based on the optimization results, the proposed VI probe is fabricated and calibrated for the root-mean-square (RMS) voltage and current with a high-voltage probe and a vector network analyzer. Then, it is evaluated through a comparison with a commercial VI probe, with the results demonstrating that the fabricated VI probe achieved a slightly higher linearity than the commercial probe: R 2 of 0.9967 and 0.9938 for RMS voltage and current, respectively. The proposed VI probe is believed to be applicable to plasma diagnostics as well as process monitoring with higher accuracy. [ABSTRACT FROM AUTHOR]

The modular multilevel converter (MMC) has become one of the most attractive converters for high-power applications such as high-voltage dc (HVDC) converters, but also fusion devices’ power supplies. This converter, thanks to the discrete-leveled output voltage and its identical submodules (SMs) by which it is composed, represents a promising alternative to replace the flywheel generator (FG) that actually provides electrical power to Axially Symmetric Divertor Experiment (ASDEX) Upgrade device’s toroidal field (TF) coils. Due to the pulsed dc operation of these coils and their high-power needs (up to 150 MW) for each experiment, a small-scale version of MMC is under development with some differences compared with conventional ones: SM’s capacitors have been replaced with supercapacitor (SC) modules to increase the amount of available stored energy, while SMs belonging to different “rows” are interconnected to simplify their control and increase the reliability of the converter. This article shows the development and the operation of a single IGBT full-bridge (FB) SM highlighting advantages of this configuration and challenges of a full-scale converter: a proper power stage filter has been designed to protect SCs from voltage spikes and high-frequency current content generated during IGBT switching events, while SM’s control has been realized, thanks to the EtherCAT protocol, which ensures synchronization in case of SMs’ parallel operation and reliability of the future full-scale converter. [ABSTRACT FROM AUTHOR]

In this article, we suggest and study a novel omnidirectional wireless power transfer (WPT) system based on unusual double-toroidal helix coils. The suggested antenna is a double-helix coil wounded around a torus. It creates two mutually orthogonal magnetic fields. One is toroidal and confined inside the torus, and another one is poloidal and is distributed outside it. The intensities of these two fields can be independently tuned via the geometry of the turns. The coupling coefficients are tailored to ensure high efficiency regardless of the receiver position and orientation, without active circuitry for tuning or control. This topology provides self-tuning functionality which is achieved when the coupling between the transmitter and receiver is minimized while the coupling in the pairs transmitter–repeater and repeater–receiver is strong. The system consists of three transmitter–repeater pairs orthogonal to each other, ensuring omnidirectional functionality. Experimental results confirm self-tunability and high efficiency for any position and orientation of receivers. The proposed double-helix antenna is applicable to different WPT applications where the coupling/decoupling between coils can be freely controlled. [ABSTRACT FROM AUTHOR]

We study numerically the possibility of programmable self-assembly of various thin-shell architectures. They include clusters isomorphic to fullerenes C20 and C60, finite and infinite sheets, tube-shaped and toroidal mesostructures. Our approach is based on the recently introduced directionally functionalized nanoparticle platform, for which we employ a hybrid technique of Brownian dynamics with stochastic bond formation. By combining a number of strategies, we were able to achieve a nearperfect yield of the desired structures with a reduced "alphabet" of building blocks. Among those strategies are the following: the use of bending rigidity of the interparticle bond as a control parameter, programming the morphology with a seed architecture, use of chirality-preserving symmetries for reduction of the particle alphabet, and the hierarchic approach. [ABSTRACT FROM AUTHOR]

The affordable, robust, compact (ARC) reactor is a tokamak fusion reactor concept currently under development by Commonwealth Fusion Systems and Massachusetts Institute of Technology. There are three important neutronics considerations for the operation of the ARC reactor: (1) breeding of enough tritium in the blanket to sustain the Dâ€"T reaction in the plasma; (2) ensuring low fluence on the superconducting toroidal field coils; and (3) assessing neutron volumetric heating in structural components. This work aims to perform a validation of the neutronics analysis approach by code-to-code comparison. State-of-the-art software stacks are employed for the neutronics analysis of the ARC reactor, and a computer-aided design (CAD) model is used directly for Monte Carlo (MC) neutron transport calculations. Three software stacks, Attila-MCNP, OpenMC-DAGMC, and Shift-DAGMC, are used to perform neutronic analyses of a 90° sector CAD model of the ARC reactor. Results show that the flux tallies calculated by the three software stacks are very close. Also, the volumetric heating and tritium breeding values have less than 0.6% relative difference between codes. [ABSTRACT FROM AUTHOR]

• A field programmable gate array (FPGA) based trigger system has been designed and developed for metallic tokamak-1 (MT-1). • The trigger system has concurrent processing capability to trigger all channel from a time reference zero. • Trigger unit has a LabVIEW based simple and user friendly graphical user interface. • Indigenously developed peripheral electronic system and field trigger modules for the safety of FGPA and computer against tokamak high voltage shocks and magnetic noise. • The trigger unit is very compact and cost effective. A field-programmable gate array (FPGA)-based trigger system (FTS) is designed and developed to replace the existing microcontroller-based system for the automation of metallic tokamak (MT-I). MT-I consists of three main systems: MT-I vacuum vessel and gas filling system, a pulsed power supply system, an electronic system based on diagnostics and data acquisition (DAQ) systems. The power supply system provides pulsed power input to the central solenoid (CS), poloidal field coils (PF), toroidal field coils (TF), and microwave systems using thyristors and insulated-gate bipolar transistors (IGBTs). The (DAQ) system acquires experimental data from MT-I diagnostics using national instruments (NI) DAQ cards. A concurrent processing system is required to incorporate time delay in the triggering process of central solenoid (CS), poloidal field coils (PF), toroidal field coils (TF), microwave source and DAQ systems. Therefore, an FTS is designed and developed to complement the processing capability, unlike a microcontroller. The FTS has twenty concurrent triggering channels, adjustable from time reference zero, and control from a simple graphical user interface (GUI) designed on LabVIEW. For current buffering and optical isolation against high voltages in the MT-I power supply system, a peripheral electronic system (PES) and field trigger modules (FTM) have been developed as part of FTS. The designed and developed FTS was tested successfully on MT-I. [ABSTRACT FROM AUTHOR]

This paper presents a novel joint-less toroidal magnet made of second-generation high-temperature superconducting (2G-HTS) tapes. This approach effectively resolves the closed-loop issue for 2G-HTS magnets and has the potential to provide higher and more stable magnetic fields. Compared with traditional 2G-HTS toroidal magnets, while the current loops of the joint-less magnets have no resistance, a decrease in magnetic field still occurs, especially in coils with insulation. Therefore, this work focuses on the decrease of the magnetic field of this novel magnet using experimental and microanalytical methods. For the first time, it was verified that, by winding two coil groups, insulated and no-insulated, parallel charging does not cause interference between them. Furthermore, the magnetic field area was expanded by finite element analysis, and simulations showed that the magnetic field converged with increasing number of coils. Besides, we found that the decrease of the magnetic field was related to the damage of the tape during slitting and winding, where insulated coils were more susceptible to damage during winding. The damage usually occurred at the starting point of the tape slitting, because the copper layer was separated due to adhesion during the winding of insulated coils, which further caused the superconducting layer to detach, resulting in a decrease in the critical current. From our perspective, benefiting from the high critical field of the 2G-HTS tapes, this novel toroidal coil structure has significant implications for the construction of compact toroidal magnets. [ABSTRACT FROM AUTHOR]

• 'PF1, 6 and CS lifting Frame' and 'Sector lifting Tool' are designed and manufactured by Korea Domestic Agency (KODA). • The lifting tools were designed considering 2 times payload according to EN standard. • The structural integrity of lifting tools was verified by load test considering 1.5 times payload at least over 15 min. • The payloads of 1566 tones and 2028 tones were applied using hydraulic cylinders and customized jig structure. • All lifting tools were successfully withstood under required test load without defect or plastic deformation. Korea Domestic Agency (ITER Korea, KODA) has procured 44 kind's assembly tools for ITER project. The assembly tools are classified by the steel structure, lifting table and lifting accessory. 'PF 1, 6 and Central Solenoid Lifting Frame' (CS Lifting Frame) and 'Sector Lifting Tool' (SLT) are the representative assembly tool classified the lifting accessory. The assembly tools handle the heavy component which over 1,000 tones for payload. In this reason, the verification of the strength of the tools is very important and the integrity of lifting accessory tools was verified by not only structural analysis but also load test at factory under 1.5 times the payload for minimum 15 min. It is the requirements of the technical specification. The main role of PF 1, 6 and CS Lifting Frame is to lift and transfer the ITER components (PF1 coil, PF6 coil and CS coil) from the assembly hall to the tokamak fit and their temporary support positions. The CS coil is heaviest component in payloads handled by CS Lifting Frame. The weight is about 1,000 tones and CS Lifting Frame was verified by 1,500 tones load test. Sector Lifting Frame is to lift and transfer the 40° sector module from Sector Sub-Assembly Tool (SSAT) to tokamak fit. The sector module consists of Vacuum Vessel (VV), VV Thermal Shield and Two Toroidal Flied Coils. The total weight of the sector module is about 1,300 tones and Sector Lifting Frame was verified by 2,000 tones load test. The load test was carried out according to test procedure approved by IO. This paper describes the overall test method, customized test JIG and result of the lifting tools for handling heavy components of ITER. [ABSTRACT FROM AUTHOR]

A 16 Tesla, large bore superconducting magnet for testing the superconducting Cable-in-Conduit Conductor (CICC) is being designed in parallel with the design of the steady state Korea fusion demonstration reactor (K-DEMO) magnet system. The test facility magnet, named SUperConducting Conductor Experiment (SUCCEX), was designed in 2014 and some design modifications of the magnet have been conducted since 2020. The peak magnetic field of the K-DEMO toroidal field coil is about 16 T, and thus the required background field of the conductor test magnet is expected to be over 15 T in a large bore of 600 mm diameter. The background field will be in addition to the K-DEMO conductor sample's self-field. To reach the target magnetic field, high current density Nb3Sn strands (Jc > 1100 A/mm2 at 4.2 K, 16 T) were employed in the design of the high field CICC. The modified SUCCEX magnet consists of concentric solenoids, with a high field inner coil (IC) and low field outer coil (OC). They are connected in series with each other, and the operating current is about 24 kA. In this study, the overall design concept is presented in relation to the results of electro-magnetic, structural analyses. [ABSTRACT FROM AUTHOR]

The Fast Discharge Units (FDUs) of the Superconducting (SC) Toroidal Field (TF) coils in the European demonstration fusion power plant DEMO warrant the machine integrity over its full lifetime against severe failure events, such as SC coil quenches or any other plant events requiring the safe TF magnet system discharge. A low (75 kA) and a high current (105 kA) configuration are under study for the TF coils for DEMO. The FDUs must be extremely reliable for the purpose of commutating in short time (∼1 s) the currents and to discharge the TF magnets safely into resistors outside of the tokamak building. Malfunctions of the FDUs must be avoided. The FDUs are considered as Safety Important Class (SIC) components that need to discharge high amounts of energy of about 161 GJ (@75 kA) resp. 118 GJ (@105 kA) stored in the DEMO TF coils. The TF FDUs Circuit Breakers (CBs) shall be installed in the lower level of the tokamak to minimize the length of the connecting busbars. The FDUs integration is challenging because of the high neutron and gamma radiation and stray magnetic fields of the tokamak. Since in DEMO the neutron fluence over lifetime is much higher than in ITER, the problems of using FDUs with electronic subsystems was expected to be more severe, so that their integration has been considered from the beginning of the DEMO project. Sufficient shielding or possible re-positioning of the whole FDUs or sensitive FDU components compared to ITER are being investigated, to reduce the neutron fluxes and neutron and gamma ray fluences. Alternative concepts, e.g., fully mechanical CBs are studied in the EUROfusion Work Package Plant Electrical System (WPPES) in parallel. This paper presents the CAD integration work on the DEMO TF FDUs supported by neutronics assessments. It is assumed the same FDU technology as in ITER. The magnet feeder´s integration is commenced at the same time. [ABSTRACT FROM AUTHOR]

In the fight against climate change through the progress of fusion technology, the Affordable, Robust and Compact (ARC) fusion reactor design is under development at Commonwealth Fusion Systems (CFS) and Massachusetts Institute of Technology (MIT)- Plasma Science and Fusion Center, in collaboration with the Ente Nazionale Idrocarburi (ENI) S.p.A. The reactor features demountable superconducting toroidal field coils and a replaceable vacuum vessel immersed in a Fluorine Lithium Beryllium (FLiBe) molten salt blanket. The low-pressure molten salt cools the divertors and the blanket, shields the magnets and acts as tritium breeder and tritium carrier. Tritium must be recovered efficiently from the salt to fuel the reactor, minimizing its inventory and its migration outside the FLiBe loop. The current work is aimed at presenting a preliminary design of a Double-Wall Heat eXchanger (DWHX) to be installed in an ARC-class reactor primary cooling system. The main feature is the presence of a flowing sweep gas gap which captures tritium and prevents its diffusion toward the secondary system. For the DWHX, thermal-hydraulic and tritium transport analyses were performed, considering a supercritical water Rankine power conversion system on the secondary side to carry out the calculations. To assess tritium extraction capabilities of the DWHX, the tritium extraction efficiency has been evaluated. Calculation results highlighted how the current configuration is not adequate to remove tritium, in fact the tritium extraction efficiency is very low. On the other side, it seems to act adequately as tritium permeation barrier. [ABSTRACT FROM AUTHOR]

The STEP (Spherical Tokamak for Energy Production) Programme aims to deliver a UK prototype fusion energy plant, targeting 2040, and a path to commercial viability of fusion. To deliver on this aim, we have performed initial scoping to identify the design point for a spherical tokamak prototype powerplant producing at least 100 MW e of net electric output and that is self-sufficient in tritium production. We have developed a rapid workflow that starts with the systems code process to develop an initial design point that can then be iterated with the plasma codes jetto and fiesta to add fidelity. Promising solutions can then be taken forward for further analysis of the tritium breeding ratio, inboard build, exhaust and power balance. Solutions that are best matched to the requirements are then taken on by a wider team to assess. We have applied this technique using a number of different input technology options and assumptions and arrived at the most promising design point for the STEP Prototype Powerplant (SPP). The unique features of this device are no inboard breeding blanket, the use of remountable magnet joints, high β and bootstrap plasma operation, a Super-X divertor, the use of just microwave heating, and enclosing poloidal field coils inside the toroidal field cage. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

High-temperature superconducting (HTS) magnets offer a novel pathway to magnetically confined fusion energy generation. Reactor-scale fusion devices require tens of kA of current to generate large toroidal fields for magnetic confinement. Operating HTS magnets at kA-current levels is difficult due to unavoidable joint resistances that require active driving of current. The large footprint and energy demand of a solid-state power supply and multi-kA current leads present significant constraints on the size, cost and efficiency of a fusion tokamak. An alternative approach is to use an HTS flux pump, which can generate and sustain persistent currents in HTS coils at the kA-levels required by fusion magnet systems. Here, we outline the design of a 60 kA transformer-rectifier flux pump for energizing and sustaining an 80 mH HTS fusion magnet. The flux pump is driven by an iron-core transformer with copper primary and HTS secondary windings. This is rectified by switching elements in the superconducting circuit into a uni-directional dc voltage output which charges the inductive load. To charge the load coil to 60 kA in 24 hours, an output dc voltage of 55 mV is needed. However, only 6 mV is required to maintain a long-term steady current of 60 kA. Such a device can significantly reduce the cost and footprint of future fusion reactor designs. [ABSTRACT FROM AUTHOR]

The three‐phase permanent magnet synchronous motor (three‐phase PMSM) has gained much popularity in many applications due to its various advantages. In addition, due to the good fault tolerance, simple inverter circuit structure and high torque density, the six‐phase PMSM has better prospects than three‐phase PMSM in biomedical devices, electric vehicles and other fields with high reliability requirements. In this study, a six‐phase axial‐flux PMSM inherently without cogging torque is proposed to improve torque density, which equips the 60° phase‐belt toroidal winding (60°TW) and slotless stator core. Based on the same effective volume, permanent magnet material, power grade and the other major motor parameters, two six‐phase PMSMs with 30° phase‐belt toroidal winding (30°TW) and 60°TW are designed, respectively. The armature reaction field, air gap magnetic density, back‐electromotive force, output torque, loss characteristic and torque‐speed characteristic are compared by using three‐dimensional (3D) finite‐element models in detail. Finally, the experimental test is carried out using the prototype of three‐phase PMSM with 120° phase‐belt toroidal winding. The experimental results indirectly verify the feasibility of 60° phase‐belt toroidal winding and correctness of 3D FEM. The 3D FEM analysis results are presented to show that torque density of the proposed six‐phase PMSM with 60°TW is improved in comparison to the six‐phase PMSM with 30°TW. Index Terms—60° phase‐belt toroidal winding, axial flux, six‐phase permanent magnet synchronous motor, torque density [ABSTRACT FROM AUTHOR]

Accurate understanding of losses and frequency‐dependent winding parameters have been an important aspect for selecting the right configuration of stranded conductors in power‐electronic inductors. An approach for modelling frequency‐dependent parameters of a winding with twisted wire bundles in toroidal inductors using a multi‐axial sliced finite element (FE) modelling approach is presented here. A 2D magnetodynamic FE problem is solved in several axial and radial slices of the inductor, accounting for the twisted conductor bundles by varying the conductor positions in the slices. Case studies are presented for different levels and pitch lengths of twisting. The approach is validated against 3D FE simulations in the case of 3–4 parallel strands and against measurements in the case of 75, 105 and 125 strands, which would be impossibly heavy for conventional 3D FE tools. The results provide insight into the effect of strand grouping, twisting levels and twisting pitch on the frequency‐dependent resistance of windings. [ABSTRACT FROM AUTHOR]

At the present time, the construction of the T-15MD Tokamak is being completed at the National Research Center Kurchatov Institute. The magnet system of the Tokamak T-15MD should provide generation and confinement of the hot plasma in the divertor configuration. The design plasma parameters are as follows: major radius of 1.48 m, minor radius of 0.67 m, elongation of 1.7–1.9 and triangularity of 0.3–0.4, plasma current of 2 MA, toroidal magnetic field on the plasma axis of 2 T. The electromagnetic system includes toroidal and poloidal coils. The installation will be equipped with the auxiliary plasma heating and current drive in quasi-stationary discharges lasting hundreds of seconds. Scenarios of quasi-stationary plasma discharges are determined by the capabilities of the electromagnetic system to create and maintain an equilibrium magnetic configuration of the plasma column. Trouble-free operation of the power supply is provided by the protection and blocking system. This work is devoted to calculations of the maximum duration of currents in toroidal and poloidal windings in order to avoid their overheating, as well as the duration and scenarios of quasi-stationary plasma discharges. [ABSTRACT FROM AUTHOR]

Results of calculation of nuclear heat deposition rate in the vacuum vessel and magnetic coils of the tokamak with reactor technologies are presented. Results obtained for D–D and D–T plasmas are compared with each other. Based on the results of the calculations, a conclusion is reached that the nuclear heat rate for the D–D plasma falls into an acceptable range below 1 mW/cm3, and the proposed vacuum vessel design is satisfactory from the point of view of radiation shielding. For the D–T plasma, the question regarding additional protection measures or limitation of discharge time remains open for discussion. Based on calculated spatial and energy distributions of the neutron field, we draw the conclusion that an increase in the thickness of protective water layer has the largest effect from the point of view of increasing radiation shielding. Nevertheless, the thickness of the water layer has to be more than doubled in the vacuum vessel, at least, in the inner segment near the equatorial plane, in order to decrease the heat load on the toroidal field coil to the level allowing long discharge when the system operates with D–T plasma. The sources of energy release in toroidal coils are estimated, along with several variants of shielding. In particular, the statement that using boron enriched with 10B isotope has nearly no impact on energy release is substantiated. The possibility of operation with D–D plasma is confirmed, and problems of radiation shielding when using D–T plasma are revealed. [ABSTRACT FROM AUTHOR]

The engineering and technical aspects of the creation of the electromagnetic system (EMS) developed using high-temperature superconductors (HTSCs) for the TRT project (Tokamak with Reactor Technologies) are considered. Some proposals on the design of superconducting the wire and EMS windings are presented, taking into account a number of restrictions imposed by equipment and operating conditions of the TRT. A brief comparison of the TRT EMS with a number of existing projects is given. [ABSTRACT FROM AUTHOR]

Here we present a new design of an inductor based on a circular magnetoelectric heterostructure containing a radially poled ring of piezoelectric lead zirconate–titanate (PZT) and an outer ring of amorphous ferromagnetic FeBSiC, mechanically coupled to each other, and a toroidal coil. Control of the device inductance by magnetic and electric fields is demonstrated. The magnetic field of 0–200 Oe was applied either along the axis of the ring structure, in the plane of the structure, or created by an additional coil wounded on the ring structure, while the electric field of 0–18 kV/cm was applied to the PZT ring. The change in the inductance reached 1000 and 410% at 300 Hz frequency when tuning by magnetic and electric fields, respectively. Uncertainty in the inductance tuning caused by the magnetic and piezoelectric hysteresis of the heterostructure was observed. The energy/power requirements for tuning/maintaining inductance of the device using magnetic and electric fields are compared. [ABSTRACT FROM AUTHOR]

HL-2M is a middle-sized copper conductor tokamak, which aims to operate close-burning and high-performance plasma for experimental magnetic fusion researches. The magnetic field coil system is an important part of HL-2M machine, consisting of a toroidal field (TF) coil system and a poloidal field (PF) coil system. In the operation of HL-2M, the TF coils operate in the environment of high current, strong magnetic field, and high stress. The structure, electromagnetic, electrical and thermal issues of TF coils were solved in design and construction phase. In this paper, the structure design principle of the demountable TF coils including some key components, was introduced, and then the construction of the TF coils, especially the industry processing, was established. [ABSTRACT FROM AUTHOR]

Important progress in the development of high-temperature superconductors (HTSC) of the second group made it possible to design the quasi-stationary tokamak with reactor technologies (TRT) with the high magnetic field (Bt0 = 8 T). The high magnetic field will ensure the achievement of plasma fusion regimes in the tokamak with the fusion energy gain Q > 1 at the considerably reduced size of the facility (R0 = 2.15 m, a = 0.57 m), and, consequently, at its reduced cost. TRT will be capable of operating in the quasi-stationary regimes (≥100 s) with hydrogen, helium, and deuterium plasmas (with the densities ne of up to 2 × 1020 m–3) and in the regimes with short (duration Δt < 10 s) deuterium–tritium plasma shots with the fusion energy gain Q > 1 limited by the radiation heating of toroidal coils. TRT is being designed as a plasma prototype for both the pure fusion reactor and the fusion neutron source for the hybrid (fusion–fission) reactor. The TRT missions are the development of the key fusion technologies and their integration in one facility. These technologies are as follows: the HTSC electromagnetic system operating at the extremely high magnetic fields; the metal and liquid-metal (lithium) first wall and innovative divertor; the unique advanced systems for the auxiliary plasma heating and non-inductive current drive, including the systems for atomic beam injection with energy of 0.5 MeV and power of several tens of megawatts, the electron cyclotron heating system based on the megawatt-power gyrotrons with a frequency of 230 GHz and a total power of ~10 MW, and the ion cyclotron heating system at frequencies of 60–80 MHz with a power of several megawatts; the tritium fuel cycle; the remote control technologies; the technologies for diagnostics capable of operating under the fusion reactor conditions; the technologies for maintaining quasi-stationary plasma discharges; and the technologies for the tokamak operation in the fusion ignition regime, in which the heating by alpha particles is the dominant heating mechanism at the axis of the plasma column, in the deuterium–tritium experiments limited by the radiation heating of the toroidal coils. The results are presented from the conceptual design of the basic TRT components, as well as the expected characteristics of its operation. It is shown that TRT has a wide window of working parameters suitable for studying the reactor operating regimes. The high magnetic field provides the necessary margins of the pressure, MHD stability, and plasma controllability variation. Implementation of the advanced divertor and first wall concepts, including those using the liquid-metal technologies, will provide the optimum choice of design options in order to reliably control the heat and particle fluxes under the reactor conditions. The advanced systems for the auxiliary heating and current drive will make it possible to implement both the pulsed and stationary regimes of the reactor operation. Calculations of the TRT discharge scenarios show that, for the DT mixture with equal content of components, the long discharges (with duration exceeding 100 s) can be realized with a neutron flux of more than 0.5 MW/m2 onto the wall, as well as the stationary discharges with a flux of approximately 0.2 MW/m2. Thus, TRT can be a real prototype of the fusion neutron source for the hybrid reactor. [ABSTRACT FROM AUTHOR]

China Fusion Engineering Test Reactor (CFETR) is a new test tokamak device in China, which is to bridge the scientific and technical gaps between ITER and DEMO. One of the main missions of CFETR is to achieve tritium self-sufficiency with a tritium breeding ratio (TBR) of no less than 1.05. In addition, the neutron shielding evaluation is also a significant issue, which could provide a significant reference for the water-cooled ceramic breeder (WCCB) blanket modification and improvements. In this work, a 3-D simplified neutronics model containing the WCCB blanket is developed and the neutronics performance is calculated based on the 1.5 GW fusion power. The nuclear analyses are carried out by the Monte Carlo N-particle (MCNP) transport code, including the TBR, neutron wall loading (NWL), fast neutron flux on inboard side, and nuclear heating deposition on main in-vessel components. It is found that the net TBR reaches 1.23, which meets the condition of tritium self-sufficiency. Moreover, it is found that the peak NWL of inboard and outboard occurs in the equatorial plane. The inboard peak NWL is 1.89 MW/m2 and the outboard peak NWL is 1.97 MW/m2. In addition, the fast neutron flux and nuclear heating deposition at the inboard midplane in toroidal field coil (TFC) are obtained. According to the radiation limits of the CFETR TFC referring to ITER, the results show the nuclear responses of the TFC meet the limit criteria. [ABSTRACT FROM AUTHOR]

This article describes a new set of tools developed to improve the conception and modeling of nonsaturated ferrite-based devices such as twin toroid phase shifters. These new simulation tools benefit from a generalized permeability tensor model able to describe the permeability tensor of a ferrite sample whatever its magnetization state. This model is coupled to a homemade 3D multi-scale magnetostatic analysis program, which describes the evolution of the magnetization through the definition of a hysteresis loop in every mesh cell. These computed spectra are then integrated into 3D electromagnetic simulation software that retains the spatial variations of the ferrite properties by using freshly developed macro programming functions. This new approach allows the designers to accurately model complex ferrite devices such as twin toroid phase shifters. In particular, we demonstrated a good agreement between simulated and measured phase shifts as a function of applied current values with a predicted maximum phase shift of 0.96 times the measured value. [ABSTRACT FROM AUTHOR]

In this work, and by means of micromagnetic simulations, we study the magnetic properties of toroidal nanomagnets. The magnetization ground state for different values of the aspect ratio between the toroidal and polar radii of the nanotorus has been obtained. Besides, we have shown that the vortex and the in-plane single domain states can appear as ground states for different ranges of the aspect ratio, while a single domain state with an out-of-plane magnetization is not observed. The hysteresis curves are also obtained, evidencing the existence of two reversal modes depending on the geometry: a vortex mode and a coherent rotation. A comparison between toroidal and cylindrical nanoparticles has been performed evidencing that nanotori can accommodate a vortex as the ground state for smaller volume than cylindrical nanorings. [ABSTRACT FROM AUTHOR]