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Publisher Copyright: IEEE 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.

The National Spherical Torus eXperiment—Upgrade (NSTX-U) experienced a series of technical problems during its initial operating campaign leading to an extended outage and recovery period. The final event that necessitated shutdown was a turn-to-turn fault in one of the inner PF coils. Investigation of the coil failure and development of remedial measures has become a major focus of the recovery. It was decided that all three inner PF coil pairs that employed similar design features and were fabricated by the same manufacturer needed to be replaced. To further reduce risk, a prototyping program was initiated to qualify suppliers. Four prospective suppliers of replacement coils were mobilized to produce an identical prototype coil that was similar in characteristics to the PF1A coil. The prototype technical evaluation included a comprehensive set of electrical tests on each of the four prototype coils and before and after destructive sectioning into two halves. This article reports on the electrical test methods and results from prototype testing, along with plans for application to the production coils.

Power quality is analyzed for the poloidal field (PF) power supply system of the Experimental Advanced Superconducting Tokamak (EAST) device. It is found that the power quality is quite different w...

Joint TEXT (J-TEXT) is an iron-core tokamak. The iron core can reduce the requirement of heating power supplies, and it also has a significant effect on the poloidal magnetic field during J-TEXT discharge. However, it is difficult to simulate an iron-core tokamak discharge process for many tokamak simulation codes (TSCs), such as TSC and DINA. Thus, the influence of the iron core on the J-TEXT magnetic field is analyzed by the theoretical model and the ANSYS model and compared with experimental results. ANSYS simulation shows that the influence of the iron core on the toroidal magnetic field can be ignored, but it has a significant impact on the poloidal magnetic field. By the comparison of ANSYS simulation and experimental results, the correction coefficient of the poloidal field with and without iron core is obtained, which can be used in some non iron-core TSC.

NSTX-U is an innovative magnetic fusion device constructed at the Princeton Plasma Physics Laboratory (PPPL) and is an upgrade of the original NSTX device that operated successfully for more than ten years. During the initial operation of NSTX-U in 2016, one of the inner poloidal field (PF) coils experienced a turn fault that necessitated a complete shutdown. A precise cause could not be identified but several contributing factors became apparent. It has since been decided to replace all three inner PF upper and lower coil pairs. Prototype and replacement inner PF coils with an improved design and construction have developed that factor in findings from the root cause analysis of the failed coil. Detailed electromagnetic (EM) and transient thermal analyses were performed for both the energized state and the cool down cycle. This article summarizes the effort taken to replace inner PF coils including improvements in design, results of analysis, experience with prototype coils, and progress with the fabrication of replacement coils.

The plasma confinement of the International Tokamak Experimental Rector (ITER) is provided by the magnetic field generated by 18 toroidal field (TF) coils while 6 poloidal field (PF) and 6 central solenoid coils have the function to drive, shape and pre-heat the plasma. Fusion for Energy (F4E), the European Domestic Agency for ITER, is responsible for the supply of 10 TF coils and 5 PF coils to the ITER project. The ITER Organization (IO) team is instead responsible for the design of such coils as well for the coordination of the activities of the different Domestic Agencies (DAs) producing the different components, and their assembly into the Tokamak. The PF coils utilize NbTi Cable-in-Conduit-Conductor and have different diameters between 8 and 24 meters and weights of up to 400 tons. Regarding the PF coils produced by F4E, so far one has been completed by the Institute of Plasma Physics Chinese Academy of Sciences (ASIPP) under a collaboration agreement with F4E. The other 4 PF coils are being produced at the ITER site in Saint Paul lez Durance, France, under F4E supervision. The first of these (PF5) will be completed by July 2020 while the last coil (PF3) will be ready be the end of 2023. The TF coils utilize Nb3Sn conductor and are manufactured with the “Wind, React & Transfer” method. The first TF coil is close to completion and will be delivered to the ITER site in early 2020. Subsequent TF coils will follow at a rate of about one every 3-4 months. In this article we will report on the production status of both PF and TF coils and, in particular, the different manufacturing strategies employed. The main challenges faced so far and the results obtained are also described.

The International Thermonuclear Experimental Reactor poloidal field (PF) coils are wound into double pancakes with NbTi cable-in-conduit conductors, which are connected by joints in shaking hands lap-type configuration. The coils are operating in pulsed mode with a maximum operating current of 55 kA and peak magnetic field of 6.4 T, utilizing electromagnetic load on the conductors and joints. A series of PF qualification joint samples modified in praying hands configuration is measured in the SULTAN facility. For some samples, a nonlinear voltage-current (VI) characteristic is observed during the assessment of joint resistance. The growth of joint resistance versus the B× I product is larger than what is expected from the magneto-resistant copper contribution. Two non-homogeneous contact resistance models are developed and combined to quantitatively evaluate the reason for the nonlinear VI behavior in combination with the relevant power dissipation and current redistribution in the joint. The simulations reveal that, for the particular pre-qualification PFJEU2 sample with resistance variation up to 3.5 nΩ, the most probable reason for the nonlinear VI characteristic is a widely spread defective connection between copper sole and shim. The electromagnetic force involves a separation effect on the mechanically and electrically weakly connected parts, resulting into a varying resistance depending on transport current and background field. The hypothesis and models are validated by an experiment on a similar sample PFJEU3 and a post-mortem examination of the PFJEU2 sample.

The inter-strand contact resistance and AC losses were measured on an International Thermonuclear Experimental Reactor (ITER) poloidal field (PF) coil joint in a parallel applied AC magnetic field. In addition, the hysteresis loss was measured as a function of the angle with the applied magnetic field on a niobium-titanium (NbTi) strand of the same type as in the joint with a vibrating sample magnetometer. The AC loss measurements were performed at four applied field conditions for combinations of 0 or 1 T offset field and 0.2 or 0.4 T sinusoidal amplitude. The hysteresis loss of the joint was compared with the measured AC loss density of the NbTi strand for the same field conditions as the joint AC loss measurement but with varying the angle of the applied field. The subsequent cable twist angles affect the hysteresis loss since the critical current and penetration field depend on the angle of the applied field. It is found that 15.5° is an effective angle for the calculation of the hysteresis loss of joint when compared to the single strand measurement. The inter-strand contact resistance measurements cover all the typical strand combinations from the five cabling stages of the individual conductors, as well as the strand combinations across the two conductors to characterize the inter-strand including the copper sole resistivity. It is the first time to measure the contact resistances and AC losses of the full-size ITER PF joint. By comparing the measured and simulated data in the JackPot-ACDC model, it is also the first time to obtain the accurate inter-strand, inter-petal and strand to copper sole contact resistivities, which are the main input parameters for the further quantitative numerical analysis of the PF joints, in any current and magnetic field conditions.