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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 55 records · Page 3
All sourcesNASA NTRSDOE OSTI
NASA NTRS1986

Sizing an emergency venting system for a cryogenic dewar

If the vacuum vessel that insulates a cryogenic dewar for a spaceborne experiment prior to launch is damaged, air will leak into the vacuum insulation space. As the sudden heat load causes the pressure to rise in the dewar, a safety disk in the emergency vent line will burst at the design pressure differential to allow vaporized cryogenic fluid to escape. The emergency vent line should be sized such that sufficient gaseous cryogen will be vented to keep the pressure inside the dewar below the design limit. On the other hand, the line should not be so large as to impose an unnecessary heat load on the dewar filled with cryogenic fluid. A vent-line computer program was generated to compute the maximum flow rate allowed for a proposed vent-line system. Parametric studies have been carried out for different burst disk pressure differentials, liquid cryogen ullage, and vent-line sizes.

Liu, C. K.↗
DOE OSTI2020

Wisconsin In Situ Penning (WISP) gauge: A versatile neutral pressure gauge to measure partial pressures in strong magnetic fields

A new type of in-vessel Penning gauge, the Wisconsin In Situ Penning (WISP) gauge, has been developed and successfully operated in the Wendelstein 7-X (W7-X) island divertor baffle and vacuum vessel. The capacity of the quantitative measurements of the neutral reservoir for light impurities, in particular, helium, is important for tokamaks as well as stellarator divertors in order to avoid fuel dilution and radiative energy loss. Penning gauges assisted by spectroscopy are a powerful tool to obtain the total neutral pressure as well as fractional neutral pressures of specific impurities. The WISP gauge is a miniaturized Penning gauge arrangement, which exploits the ambient magnetic field of magnetic confinement fusion experiments to establish the Penning discharge. Then, in situ spectroscopy is conducted to separate the fractional neutral pressures of hydrogen, helium, and possibly also other impurities. The WISP probe head was qualified using the magnetic field of the Magnetized Dusty Plasma Experiment at Auburn University between 0.25 T and 3.5 T [E. Thomas et al ., J. Plasma Phys. 81, 345810206 (2015)]. The in-depth quantitative evaluation for hydrogen and helium will be shown as well as an exploration of nitrogen, argon, and neon. A power law scaling between current I and pressure p, I = f (Gas,V) p n⁽Gas, B⁾ , was shown. The factor f is gas and anode potential dependent, while n is gas and magnetic field strength dependent. Pressure measurements from 0.1 mbar and down to 1 × 10 -5 mbar were achieved, demonstrating a reliable operating range for relevant pressure levels in the divertor and main vessel regions in current and future fusion devices, with a time resolution of up to 1 kHz. The lowest achievable pressure measurement increases with an increase in B and can be shifted with the anode potential V . At W7-X, the WISP probe head was mounted on an immersion tube setup that passes through the cryostat and places the probe head close to the plasma. Two probe heads were positioned in different divertor pump gaps, top and bottom, and one close to the plasma on the midplane in one module. The gauges were in situ calibrated together with the ASDEX pressure gauges [G. Haas and H.-S. Bosch, Vacuum 51, 39 (1998)]. Data were taken during the entire operation phase 1.2b, and measurements were coherent with other neutral gas pressure gauges. For the spectroscopic partial pressure measurements, channels of a spectroscopic detection system based on photo-multipliers, a so-called filterscope [R. J. Colchin et al ., Rev. Sci. Instrum. 74, 2068 (2003)], provided by the Oak Ridge National Lab were used.

47 OTHER INSTRUMENTATION↗
DOE OSTI2022

Improvements to the High-Field-Side Transient CHI System on QUEST

Transient coaxial helicity injection (CHI) is a promising candidate for solenoid-free plasma current start-up in a low-aspect-ratio tokamak in support of developing fully non-inductive scenarios. The aim of the transient CHI research on QUEST is to develop a reactor-relevant CHI design. On QUEST, a CHI discharge is initiated by driving current along magnetic field lines that connect an electrically insulated electrode plate (which is referred to as a bias electrode) to a vessel component at the bottom region of the spherical tokamak. In the first application of CHI on QUEST, the electrically insulated electrode plate was biased with respect to the outer vessel in a configuration referred to as low-field-side (LFS) injection. To maintain a narrow injector flux footprint width throughout the discharge, QUEST is now developing a high-field-side (HFS) injection configuration, in which the electrically insulated electrode plate is biased with respect to the inner vessel components. Through the implementation of a CHI-dedicated gas injection system, studies in the HFS injection configuration have now demonstrated good magnetic flux evolution into the vacuum vessel. Toroidal currents up to 43 kA have been achieved, and the generated current has increased in proportion to the magnetic flux connecting the electrodes. Finally, these results which show agreement with the CHI-scaling suggest that much higher levels of toroidal current can be generated on QUEST in an optimized CHI system in which the magnetic flux connecting the CHI electrodes is further increased.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
NASA NTRS1992

100-kWe lunar/Mars surface power utilizing the SP-100 reactor with dynamic conversion

Results are presented from a study of the coupling of an SP-100 nuclear reactor with either a Stirling or Brayton power system, at the 100 kWe level, for a power generating system suitable for operation in the lunar and Martian surface environments. In the lunar environment, the reactor and primary coolant loop would be contained in a guard vessel to protect from a loss of primary loop containment. For Mars, all refractory components, including the reactor, coolant, and power conversion components will be contained in a vacuum vessel for protection against the CO2 environment.

Harty, Richard B.↗
DOE OSTI2026

A MINi Powder INjector (MINPIN) for the SMART tokamak

Here, this article presents the design and performance characterization of a MINi Powder INjector (MINPIN) developed for boron wall conditioning (boronization) in the SMall Aspect Ratio Tokamak (SMART) at the University of Seville, Spain. SMART studies positive and negative triangularity plasmas within the low aspect ratio range of 1.4 < R/a < 3, with a major radius (R) of 0.40–0.60 m and a minor radius (a) of 0.20–0.33 m. MINPIN, in design, will gravitationally deliver boron (B) powder into the plasma, which ablates and transports the injected material. Mounted on the top of the vessel, the MINPIN will drop controlled amounts of powder using a piezoelectric blade driven by a signal generator that vibrates at a set frequency. The B will then be deposited in a layer on SMART’s stainless steel vacuum vessel walls, and this boronization process reduces the influx of intrinsic wall impurities, such as oxygen, carbon, nitrogen, and iron, which radiate power and cool the plasma. In the laboratory, we calibrated the drop rate for 1–10 mg/s, which will enable controlled boronization coating while achieving efficient powder ablation. We propose MINPIN’s operational scenarios based on the calibrated drop rate and estimated drop duration, assuming the distance from MINPIN to SMART’s plasmas.

Boron powder↗
DOE OSTI2021

Application of motional Stark effect in situ background correction to a superconducting tokamak

A polychrometer-type motional Stark effect (MSE) diagnostic technique, originally developed for the Alcator C-Mod tokamak, has been extended and applied to the Korea Superconducting Advanced Tokamak Research (KSTAR) device, the long-pulse superconducting tokamak, for the first time. It demonstrates a successful in situ subtraction of the polarized reflections off the vacuum vessel wall, sometimes up to half the total signal in some sightlines. To avoid the secondary neutral beam emission that may contaminate conventional beam-into-gas calibrations, a new approach, where the beam-into-gas measurements are made at various torus pressures with fixed vacuum fields, has been devised, which is possible with the stable superconducting coil systems of KSTAR. The validity of this new calibration scheme has been checked via plasma jog experiments. The experimental evidence of the polarized background light and the necessity of its correction in the MSE measurements made in KSTAR are presented as well.

47 OTHER INSTRUMENTATION↗
DOE OSTI2024

Boundary condition effects on runaway electron mitigation coil modeling for the SPARC and DIII-D tokamaks

Abstract Extended-MHD modeling of planned Runaway Electron Mitigation Coils (REMC) for SPARC and DIII-D is performed with the NIMROD code. A coil has been designed for each machine, with the two differing in shape and location, but both having n = 1 symmetry (with n the toroidal mode number). Compared to previous modeling efforts, three improvements are made to the simulations boundary conditions. First a resistive wall model is used in place of an ideal wall. Second, the ThinCurr code is used to compute the time-dependent 3D fields used as magnetic boundary conditions for the simulations. Third, the simulation boundary is moved from the first-wall location to the Vacuum Vessel (VV), which extends the boundary past the location of the internal REMC. To remove the 3D coil from the simulation domain, an equivalent set of 3D fields is calculated at the VV boundary that produce approximately the same field distribution at the last closed flux surface assuming vacuum between the two. Each of these three boundary condition improvements leads to an improvement in the predicted performance of the REMC for both machines. The resistive wall alone primarily effects the resonance of the coil with the plasma after the TQ, affecting the q-profile evolution in the SPARC modeling, and allowing the applied spectrum to be modified in response to the plasma in the DIII-D modeling. The movement of the simulation boundary has the most significant effect on the RE confinement overall, including in the early stages, particularly for a DIII-D inner wall limited equilibrium, where the RE loss fraction increases from 90% to > 99%, with SPARC RE losses also occurring much earlier when the boundary is placed at the VV.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
NASA NTRS1996

Ultralow Friction in a Superconducting Magnetic Bearing

Passive levitation by superconducting magnetic bearings can be utilized in flywheels for energy storage. Basic design criteria of such a bearing are high levitation force, sufficient vertical and horizontal stability and low friction. A test facility was built for the measurement and evaluation of friction in a superconducting magnetic bearing as a function of operating temperature and pressure in the vacuum vessel. The bearing consists of a commercial disk shaped magnet levitated above single grain, melt-textured YBCO high-temperature superconductor material. The superconductor was conduction cooled by an integrated AEG tactical cryocooler. The temperature could be varied from 50 K to 80 K. The pressure in the vacuum chamber was varied from 1 bar to 10(exp -5) mbar. At the lowest pressure setting, the drag torque shows a linear frequency dependence over the entire range investigated (0 less than f less than 40 Hz). Magnetic friction, the frequency independent contribution, is very low. The frequency dependent drag torque is generated by molecular friction from molecule-surface collisions and by eddy currents. Given the specific geometry of the set-up and gas pressure, the molecular drag torque can be estimated. At a speed of 40 Hz, the coefficient of friction (drag-to-lift ratio) was measured to be mu = 1.6 x 10(exp -7) at 10(exp -5) mbar and T = 60 K. This is equivalent to a drag torque of 7.6 x 10(exp -10) Nm. Magnetic friction causes approx. 1% of the total losses. Molecular friction accounts for about 13% of the frequency dependent drag torque, the remaining 87% being due to eddy currents and losses from rotor unbalance. The specific energy loss is only 0.3% per hour.

Bornemann, Hans J.↗
DOE OSTI2022

Bus Design for the Poloidal Field Coils of the NSTX-Upgrade Fusion Device

The National Spherical Torus eXperiment (NSTX) has undergone a major upgrade to NSTX-U at Princeton Plasma Physics Laboratory (PPPL). NSTX upgrade (NSTX-U) will double the toroidal field, plasma current, and neutral beam injection heating power, as well as significantly increase the pulse duration. NSTX-U uses three poloidal field (PF) coils at the vessel top and three at the bottom near the divertor areas to control the local plasma shape there. These shaping coils operate at a maximum terminal voltage of 2 kV, corresponding to a maximum current about 20 kA. To supply the power to the PF coils, hard copper bus bars are typically used near the vacuum vessel, while water cooled flexible cables are used away from the vessel. The detailed design of the hard copper bus bars will be covered in this article. During operation, these hard bus bars are subject to high electromagnetic (EM) forces, thermal displacement loads, as well as plasma halo loads due to disruption. The EM, thermal, and structural analysis were performed, and the results revealed that, with the worst loads combined, the new design meets the NSTX-U thermal, structural, and fatigue cycle requirements. Finally, the manufacturing and installation process of the bus work will be discussed at the end of this article.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Neutronics Analysis of Shielding Material Alternative to Tungsten Carbide in the FESS-FNSF Facility

Neutron transport calculations have been performed on advanced shielding materials. Metal hydrides and borohydrides were evaluated to find an alternative to tungsten carbide (WC), which is used in the in-vessel components. The study was conducted using a 22.5° sector and a detailed 360° geometry of the Fusion Energy System Studies-Fusion Nuclear Science Facility (FESS-FNSF) using OpenMC and FENDL-3.2b nuclear data library. Here, the neutronics analysis in this article was concentrated on calculating total nuclear (neutron and photon) heating at the magnet and the radiation damage of the inboard vacuum vessel (IBVV). For example, yttrium hydride (YH 2 ) and vanadium hydride (VH 2 ) showed lower radiation damage values compared to WC and other metals. Whereas alternative shielding materials did not show a significant change in the magnet nuclear heating.

36 MATERIALS SCIENCE↗
DOE OSTI2025

DESIGN ISSUES AND QUALIFICATION OF HYDROFORMED DOUBLE WALLED EXPANSION JOINTS IN VACUUM SERVICE

The Vacuum Auxiliary System for ITER is devoted to pumping out, venting, and purging the vacuum volumes of the tokamak. Over 5000 clients including the cryostat and vacuum vessel, at 8500 m3 and 1400 m3 respectively, are serviced by approximately 150 pumping stations through 6 km of pipework. The piping and functions provided to the clients include component operation, routing of potentially tritiated gases, and timely leak localization. Expansion joints are used to reduce loadings on pumps and piping stresses but also to qualify in-line components having low allowable design loads. The ITER project utilizes vacuum valves that are not designed to withstand typical piping system loadings and require detailed design and evaluation. Loading scenarios prescribed by governing system specifications must be considered. To maintain allowable design loads for components and to keep piping stresses below code requirements, double walled hydroformed expansion joints have been employed in the design. These are not standard items and require custom fabrication. Installation space constraints and adjacent pipe support attachment location availability at ITER limit standard expansion joint installation guidelines provided by the Expasion Joint Manufacturers Association (EJMA). Careful consideration must be given to the analysis model to ensure proper function and life expectancy of the component. These considerations include accurate accounting of thrust forces, thermal movements, seismic accelerations, equipment and building differential displacements. In addition to displacements, the process internal, external, and interspace pressures affect the qualification and selection of the double walled expansion joints. The calculation results shall confirm that deflections, forces, and moments are reasonable for the size and type required for the system’s demands as evaluated against the manufacturer’s design.

Clark, Forrest [ORNL] (ORCID:0009000678106843)↗
DOE OSTI2024

Realization of a gas puff imaging system on the Wendelstein 7-X stellarator

A system for studying the spatiotemporal dynamics of fluctuations in the boundary of the W7-X plasma using the “Gas-Puff Imaging” (GPI) technique has been designed, constructed, installed, and operated. This GPI system addresses a number of challenges specific to long-pulse superconducting devices, such as W7-X, including the long distance between the plasma and the vacuum vessel wall, the long distance between the plasma and diagnostic ports, the range of last closed flux surface (LCFS) locations for different magnetic configurations in W7-X, and management of heat loads on the system’s plasma-facing components. The system features a pair of “converging–diverging” nozzles for partially collimating the gas puffed locally ≈135 mm radially outboard of the plasma boundary, a pop-up turning mirror for viewing the gas puff emission from the side (which also acts as a shutter for the re-entrant vacuum window), and a high-throughput optical system that collects visible emission resulting from the interaction between the puffed gas and the plasma and directs it along a water-cooled re-entrant tube directly onto the 8 × 16 pixel detector array of the fast camera. The DEGAS 2 neutral code was used to simulate the Hα (656 nm) and HeI (587 nm) line emission expected from well-characterized gas-puffs of H2 and He and excited within typical edge plasma profiles in W7-X, thereby predicting line brightnesses used to reduce the risks associated with system sensitivity and placement of the field of view. Operation of GPI on W7-X shows excellent signal-to-noise ratios (>100 at 2 Mframes/s) over the field of view for minimally perturbing gas puffs. The GPI system provides detailed measurements of the two-dimensional (radial and poloidal) dynamics of plasma fluctuations in the W7-X edge and scrape-off layer and in and around the magnetic islands outside the LCFS that make up the island divertor configuration employed on W7-X.

Instruments & Instrumentation↗
DOE OSTI2024

On a High-Mirror Stellarator Reactor Exploratory Concept With Neutrons Concentrated on Centrifuge Liquids

In the framework of fusion energy research based on magnetic confinement, stellarators allow numerous degrees of freedom for the design of the magnetic trap and plasma shape. Taking advantage of these features, some plasma shapes might benefit several of the many integrated elements involved in commercial fusion reactors, like, e.g., decreasing the number, mass and complexity of replaced activated in-vessel components (IVC) (i.e., by using liquids), extraction of large power, tritium generation, and remote maintenance. Certainly, free-surface liquid materials were proposed for tokamaks and field-reversed configuration (FRC) to try to improve some of such elements, i.e., in advanced power extraction (APEX) studies. Some reactor-relevant quasi-isodynamic (QI) magnetic configurations exhibit a relatively straight sector of plasma and high magnetic mirror. The combination of those elements and possibilities in a single stellarator reactor concept might have some advantages, in spite of the uncertainties due to the current low technological readiness level (TRL). The proposed and studied reactor concept is based on a vacuum vessel having short curved sectors and longer wide cylindrical sectors, which encloses a high-mirror low-vertical-excursion magnetic configuration, and swirling liquids or rotating cylinders, which centrifuge molten Li salts located at the low field region. Thus, the molten salts (if possible covered by a thin layer of liquid lithium) would be located on the internal perimeter of the cylinder, to act as particle exhaust (except for helium), neutron power extractor, and tritium breeder. The high-mirror feature tries to concentrate the neutron power at the cylindrical sectors, which might avoid using breeding materials at the curved sectors. The different elements of the concept are exploratorily studied and defined, and the difficulties assessed.

centrifuge↗
DOE OSTI2022

Material Plasma Exposure eXperiment (MPEX) High Heat Flux Bellows for Component Alignment in High Microwave Environment

The Material Plasma Exposure eXperiment (MPEX) device is a linear plasma device developed to perform plasma material interaction experiments under the conditions prototypic of a fusion reactor divertor. MPEX has multiple systems that must be precisely aligned to the plasma axis, including an electron cyclotron heating system that emits up to 400 kW of microwave power into the vacuum vessel. Five distinct systems require precise alignment on the MPEX device, thus requiring four bellows, all of which are adjacent to the plasma at a relatively high heat flux of approximately 47 kW/m2 and microwave power regions. The MPEX high heat flux bellows (HHFB) is designed to deliver 6 degrees of freedom positioning. The HHFB includes titanium–zirconium–molybdenum (TZM) inserts that are brazed into a Glidcop AL-15 body using a high-temperature braze alloy, thus blocking direct line of sight to an edge-welded bellows from the plasma and microwave screen to block microwaves from the bellows. Further, a custom ConFlat knife edge is machined into the Glidcop AL-15, so the vacuum flanges do not need a braze or weld joint on the vacuum interface. Fingerstock or copper mesh is used to restrict microwave power from entering the interstitial space between the water-cooled Glidcop AL-15 body and edge-welded bellows. Glidcop AL-15 was selected as the material of choice for the water-cooled body because it can maintain mechanical integrity at elevated temperatures, and it also allows for a high-temperature braze. TZM was selected for its machinability and compatibility with vacuum and plasma requirements. A test article is also being considered for a similarly shaped component, the MPEX limiter, that will demonstrate the integrity of the braze joint under high thermal load. Results from this testing will be extrapolated to deduce the lifetime and integrity of the HHFB design.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

The kinetics of the PuO 2 to Pu 2 O 3 conversion

Here in an oxidizing environment, the oxide formed on plutonium (Pu) metal is composed of a plutonium dioxide (PuO 2 ) top layer and a thin cubic plutonium sesquioxide (Pu 2 O 3 ) middle layer. In a reducing environment, the PuO 2 layer auto-reduces to cubic Pu 2 O 3 . The speed and extent of this conversion depend on the combination of temperature and time. While PuO 2 provides a strong diffusion barrier against unwanted Pu corrosion by gaseous species (like hydrogen), Pu 2 O 3 does not, since its crystal structure has chains of oxygen vacancies. The kinetics of the PuO 2 reduction are, therefore, of fundamental interest and enable researchers to better protect Pu from corrosion. In this report, the oxygen-diffusion-limited kinetics of the dioxide to sesquioxide conversion were obtained by dynamically heating a PuO 2 -covered Pu sample from 294 to 418 K in a high-vacuum vessel equipped with an in situ spectroscopic ellipsometer. The physical/chemical constraints in the conversion process were combined with the ellipsometry method of multi-sample analysis to track the percentage of PuO 2 and to compute the extent of Pu 2 O 3 formation. The resulting diffusion coefficients were compared against and then combined with complementary literature data to produce a comprehensive set of kinetic parameters for reliably modeling oxide conversion over a larger temperature range than spanned by prior studies. The extracted thermal activation energy barrier (43.7 kJ/mol) and pre-exponential factor (5.0 × 10 -10 cm 2 /s) for the oxygen-diffusion-limited process can be used to accurately model the PuO 2 to Pu 2 O 3 transformation in vacuum and/or inert gas applications.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Structural Analysis of the US ITER ECH Transmission Line System

The electron cyclotron heating (ECH) and current drive system is one of the main plasma heating systems for ITER. It uses high-power microwave beams with the power deposition location steerable across the plasma cross section. Microwave power is conveyed via transmission lines (TLs) that run from the gyrotrons in the radio frequency building through the assembly hall and tokamak building to the ECH launchers within the tokamak vacuum vessel. The ECH system includes a vast array of interconnected TL waveguides, in-line components, and support structures.Finite element (FE) modeling provides an essential means of simulating the system, applying loads and determining deflections, rotations, forces, moments, and stresses in order to evaluate various structural and microwave transmission performance metrics. A representative FE model of the overall ECH TL system is developed in ANSYS®. This top-level model defines the centerline of the waveguide system. Waveguide segments are represented by line elements (beams and pipes) with equivalent section properties, and support structures are represented by boundary conditions. A systematic approach is used to model each ECH component with lumped masses and structurally equivalent stiffness matrices or ANSYS superelements.The top-level TL FE model is used to evaluate the various loads (thermal, vacuum, seismic, etc.) and operating scenarios. The top-level model directly calculates stresses in the straight aluminum waveguide segments. The model provides the forces and moments acting on the in-line components for detailed submodel assessments. Displacement results from the top-level analysis feed into a separate microwave performance model to help determine operational efficiency. All TL performance and thermal-structural requirements are met, as specified by the applicable codes and standards, and successfully documented in numerous technical reports and demonstrated at the final design review.

Shanmugasundaram, Aravind↗
DOE OSTI2025

Final design of outer poloidal field coils for COMPASS Upgrade

The paper aims to provide an overview of the final design of the outer poloidal field coils system for COMPASS Upgrade. Due to a relatively unique combination of design features of the new device (placement in high toroidal magnetic field, copper coils sub-cooled to liquid nitrogen temperature, hot vacuum vessel operation), the poloidal field coils design faced several interesting engineering challenges. Most notably high electromagnetic stress in the coil conductor, compatibility of the insulation to thermal and mechanical stress and design of coil supports with given spatial constraints while able to accommodate coil movements and withstand considerable mechanical stress. In conclusion, the paper describes solution to these challenges together with underlying analyses and mechanical tests.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

A new tubular ironless linear induction motor using non-interrupted winding method

The ironless linear induction motor device can be used for any application which requires a large motion span or rotation restrictedly by the ironless feature due to the strong magnetic field environment. One of the possible applications is ITER diagnostics shutters. The tokamak reactor systems have a severe radiation issue during the operation. The traditional Kapton-insulated wires have significantly lower radiation resistance than an expected large radiation dose at ITER Motion Stark Effect (MSE) shutter location. Furthermore, the traditional single-sided linear induction motor winds the wire to make the N-turn pitch coils independently and then connect them in series to make a one-phase winding. Therefore, it is a big challenge to make so many radiation-protection connectors in such a limited space as inside the vacuum vessel of fusion reactors. The approved mineral-insulated kinds of cables need a large bend radius for the end turns, so the traditional single-sided linear motor with two-layer three-phase windings need to be upgraded to the tubular linear motors. The new proposed tubular ironless linear induction motor can not only overcome the large bend radius issue with the mineral-insulated cables but also allow the motor coils wound continuously around the tubular-shaped stator, forward and backward N times to make a N-turn winding for each phase, in order to resolve the radiation protection issue with the winding connection. The 3D transient Maxwell electromagnetic model with a tubular three-phase linear induction motor has been analyzed. Furthermore, this paper will present the detailed design and winding method of ironless linear induction motor actuators, and results of the 3D transient numerical simulation of the actuator performance.

Ironless↗