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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.

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.

Disruption simulations with DINA code are performed for JT-60SA design. The simulation results have been applied for the design of many components, not only for the vacuum vessel and in-vessel components, but also for peripheral components. For instance, for the design of in-vessel coils, stabilizing plate and magnetic sensors, EM force induced by halo current and eddy current at disruption were calculated. For design of poloidal field (PF) coils, the power supply of PF coils and refrigerator system for super conducting coils, eddy current of PF coils and AC loss of superconducting coils were evaluated.

The National Spherical Torus eXperiment Upgrade (NSTX-U) is an innovative magnetic fusion device constructed at the Princeton Plasma Physics Laboratory (PPPL). In 2016, because of the failu...

In order to simulate a full tokamak discharge, a fast integrated tokamak modeling tool for scenario design, METIS, is coupled with a quasi-static free-boundary magnetic equilibrium code, FEEQS. The first code is able to compute at each time the plasma equilibrium, the current density and plasma pressure profiles. The generated outputs are iteratively used by the second code that computes the poloidal field coils currents needed in order to obtain a given plasma shape. This is used to assess the feasibility of scenarios designed by integrated modeling simulations. This approach is much faster with respect to full simulators, equipped with specific feedback controllers, and can be regarded as complementary or preparatory to much more time-consuming scenario control optimisation. The results obtained with the equilibrium code are benchmarked with data already available and computed using the TOSCA code. To illustrate how the coupled codes work, the optimization of one of the reference scenarios of the JT-60SA tokamak is carried out. The use of this new simulation tool for tokamak scenario design is discussed.

Observations of the solar magnetic cycle showed that the amplitude of the cycle did not grow all the time in the past. Thus, there must be a mechanism to halt the growth of the magnetic field in the Sun. We demonstrate a recently proposed mechanism for this under the Babcock--Leighton dynamo framework, which is believed to be the most promising paradigm for the generation of the solar magnetic field at present. This mechanism is based on the observational fact that the stronger solar cycles produce bipolar magnetic regions (BMRs) at higher latitudes and thus have higher mean latitudes than the weaker ones. We capture this effect in our three-dimensional Babcock--Leighton solar dynamo model and show that when the toroidal magnetic field tries to grow, it produces BMRs at higher latitudes. The BMRs at higher latitudes generate a less poloidal field, which consequently limits the overall growth of the magnetic field in our model. Thus, our study suggests that the latitudinal variation of BMRs is a potential mechanism for limiting the magnetic field growth in the Sun., Published in ApJL; 6 pages, including 4 figures

We explore the thermal and magnetic-field structure of a late-stage proto-neutron star. We find the dominant contribution to the entropy in different regions of the star, from which we build a simplified equation of state for the hot neutron star. With this, we numerically solve the stellar equilibrium equations to find a range of models, including magnetic fields and rotation up to Keplerian velocity. We approximate the equation of state as a barotrope, and discuss the validity of this assumption. For fixed magnetic-field strength, the induced ellipticity increases with temperature; we give quantitative formulae for this. The Keplerian velocity is considerably lower for hotter stars, which may set a de-facto maximum rotation rate for non-recycled NSs well below 1 kHz. Magnetic fields stronger than around $10^{14}$ G have qualitatively similar equilibrium states in both hot and cold neutron stars, with large-scale simple structure and the poloidal field component dominating over the toroidal one; we argue this result may be universal. We show that truncating magnetic-field solutions at low multipoles leads to serious inaccuracies, especially for models with rapid rotation or a strong toroidal-field component., 21 pages, 16 figures. Matched version accepted for publication in MNRAS. Some rewritten discussion and one entropy-profile plot changed from v1; main results unaffected

Magnetospheric ultralow‐frequency (ULF) waves contribute to space weather in the solar wind‐magnetosphere‐ionosphere system. The monitoring of these waves by space‐ and ground‐based instruments, however, produces big data, which are difficult to navigate, mine, and analyze effectively. We present sonification, the process of converting an oscillatory time series into audible sound, and citizen science, where members of the public contribute to scientific investigations, as a means to potentially help tackle these issues. Magnetometer data in the ULF range at geostationary orbit have been sonified and released to London high schools as part of exploratory projects. While this approach reduces the overall likelihood of useful results from any particular group of citizen scientists compared to typical citizen science projects, it promotes independent learning and problem solving by all participants and can result in a small number of unexpected research outcomes. We present one such example, a case study identified by a group of students of decreasing‐frequency poloidal field line resonances over multiple days found to occur during the recovery phase of a coronal mass ejection‐driven geomagnetic storm. Simultaneous plasma density measurements show that the decreasing frequencies were due to the refilling of the plasmasphere following the storm. The waves were likely generated by internal plasma processes. Further exploration of the audio revealed many similar events following other major storms; thus, they are much more common than previously thought. We therefore highlight the potential of sonification and exploratory citizen science in addressing some of the challenges facing ULF wave research.

The Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP) is responsible for the construction and performance testing of the 30 pairs of the international thermonuclear experimental reactor high-temperature superconducting (HTS) current leads. The first pair of poloidal field (PF) coil type, 52-kA HTS current lead prototypes was built and cold tested in ASIPP in mid-2016. The test results demonstrated their excellent performance on low joint resistance, long loss of flow accident time, and high current-sharing temperature. The overheating time, mass flow, and heat loads to 5-K ends also meet the requirements. This paper summarizes the major test results for the PF 52-kA HTS current lead prototypes.

This paper clearly demonstrates the architecture of time synchronization network of experimental advanced superconducting tokamak (EAST) poloidal field (PF) power supply control system and analyzes characteristics of the IEEE 1588 in detail at the same time. IEEE 1588 has unique advantages for this paper to utilize as protocol in order to improve the accuracy of the time synchronization network (TSN). Consequently, we design a time synchronization system based on PCI eXtensions for Instrumentation-6683H. This paper utilizes one master node to synchronize all slave nodes on the time network in each second. The result shows that the synchronization accuracy can be achieved within 50 ns. According to results of experiments, this system can perfectly meet the requirements of EAST PF power supply control system.

Merging compression is an efficient and robust method for plasma start-up which can achieve both high plasma current and high temperatures at fusion relevant densities. It involves the formation of plasma around two in-vessel poloidal field coils followed by a magnetic reconnection event after which the plasma can be compressed. In this paper we extrapolate from START/MAST experimental data to predict the performance of ST40 (currently under construction). In our analysis we found that induced eddy currents within passive elements greatly affected the scenario.

We calculate a set of ITER poloidal field (PF) coil currents at the initial magnetization phase using the quadratic programming algorithm. The optimization is performed by exploiting the total magnetic energy of the ITER PF system as a cost function to be minimized. We also develop a simple analytic model which can make a rapid evaluation of temperature margin of superconducting PF coils for a given set of coil currents. This calculation is necessary for a quick examination of the compatibility of the PF scenario with engineering constraints. For an easy adaptation to the future integrated ITER simulator, all modules are developed using the ITER Integrated Modeling Analysis Suite (IMAS) data structure and the Kepler framework.

The poloidal field (PF) coils are one of the CFETR magnet system which winding by CICC conductor. In order to meet the design requirements of demanding, the coil insulation shall be processed by vacuum pressure impregnation (VPI). After the resin impregnating the coil insulation, a subsequent high temperature cure is imposed for a relatively long period of time to fully cure the resin throughout the insulation. During the curing, the temperature must be controlled even on whole coil. All CFETR PF coil are fabricated by stacking several double-pancakes (DP), due to the complex structure, the common outer heating method can hardly satisfy the curing temperature requirement. So an inner heating which assisted by common outer heating (hybrid heating) method was researched. The experiment results shows under hybrid heating method, the curing temperature on the coil insulation can guaranteed even and the insulation quality can satisfy design requirement.

Fault suppression capability (FSC) test with DC short downstream DC reactor of international thermonuclear experimental reactor (ITER) poloidal field (PF) converter is performed to verify the thermal stability characteristics of the converter bridge arms. In this paper, firstly, DC short downstream DC reactor of ITER PF converter is analyzed to get the fault current. Then, the appropriate test scheme is designed and the feasibility of the test scheme is verified by simulation based on test platforms in Institute of Plasma Physics, Chinese Academy of Science (ASIPP). Finally, the effectiveness of the designed test scheme results are verified by the test platforms in ASIPP.

The PF1 coil is one of six poloidal field coils of the ITER magnet system. In accordance with the PF-1 coil procurement arrangement, JSC NIIEFA have manufactured and tested qualification samples of the PF-1 coil electrical turn insulation. Insulation qualification samples should withstand static and cyclic mechanical tests at 77 K in accordance with specified parameters[1]. In 2015, JSC NIIEFA carried out mechanical tests of the PF-1 coil insulation qualification samples. These tests demonstrated that mechanical properties of the turn insulation of PF-1 coil meet the ITER requirements. The successful test results prove that developed technologies could be used for manufacturing of the electrical insulation of the PF-1 coil. The following tests at 77 K have been carried out in 2015: tensile, interlaminar shear, shear-compression, and push-out tests in the specimens cut from 3 × 3 mock-up sample. Samples for tensile and interlaminar shear tests were cut from plates, which simulate multilayer turn insulation of PF-1 coil. Shear-compression samples were cut from the sandwich plates, which simulate bonding of the insulation to the stainless steel jacket of PF1 conductor. All types of qualification samples successfully passed mechanical tests at 77 K. This paper describes: preparation of insulation qualification samples; VPI process of the insulation qualification samples; and mechanical tests of the qualification insulation samples at 77 K.

This paper summarizes the work on the manufacturing and testing of conductor samples for ITER poloidal field 1&6 coils. The work was carried out by the IO and multiple institutions in the EU, Switzerland and Russia. Five samples were manufactured and tested in the SULTAN test facility. Analyses of dc performances, ac, and mechanical losses test results have been performed. All measurements were carried out before and after cyclic mechanical load created by the electromagnetic forces. DC test results of all samples tested demonstrated that the conductor performances are not affected by cycling mechanical load and meet the required value of current-sharing temperature T cs ≥ 5.8 K at magnetic field B max = 6.4 T and operating current I op = 48 kA. Measurements of the ac and mechanical losses were performed by gas flow calorimetric method. AC loss measurements were carried out at different operating currents and background magnetic fields. AC loss of tested samples showed no change under cycling mechanical load, but showed a correlation of ac loss versus background magnetic field. Mechanical loss, as had been expected, decreased significantly after first ten cycles then reaches a stabile value. The value of mechanical loss even in virgin condition does not affect workability of a magnet.

This paper describes the feasibility analysis of the fuzzy logic control to increase the performance of the ITER poloidal field (PF) converter systems. A fuzzy-logic-based controller is designed for ITER PF converter system, using the traditional PI controller and Fuzzy controller (FC), the dynamic behavior and transient response of the PF converter system are compared under normal operation by analysis and simulation. The analysis results show that the fuzzy logic control can achieve better operation performance than PI control.

The Poloidal Field (PF) coils are one of the main sub-systems of the ITER magnets. The PF6 coil is being manufactured by the Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP) as per the Poloidal Field coils cooperation agreement signed between ASIPP and Fusion for Energy (F4E).The ITER PF6 winding pack is composed by stacking of 9 double pancakes. Each double pancake is wound with a “two-in-hand” configuration. This paper describes the ITER PF6 double pancakes winding line, including layout and main parameters of the winding line, features of main units and the commissioning trial with PF5 empty jacket conductor.

An upgrade to the lower divertor is currently being planned for EAST superconducting tokamak, aiming at reaching over 400 s long-pulse H-mode operations with a full metal wall and a divertor heat load of ˜10 MW/m2. A new divertor concept for EAST, “Tightly Baffled Divertor”, suited to water- cooled W/Cu plasma face components (PFC) with minimized divertor volume, has been proposed to achieve Te,target

One of the most crucial issues in the design of the ITER machine is the electromagnetic (EM) loads associated with eddy currents induced in the conducting structures during plasma disruptions. The ITER database contains tens of calculation scenarios for possible plasma disruption events. The duration of these scenarios is usually limited by the end of the plasma current quench when the toroidal plasma current decreases to zero. For further EM analysis, the currents in the poloidal field coils are conventionally assumed constant after the end of the current quench. Such approach is not good for estimation of electromagnetic loads acting on the tokamak components located outside vacuum vessel. A possible way to solve this problem is to employ pre-determined coil current variations over the entire period of observation including time interval after the current quench. This paper describes an approach to such electromagnetic calculations and comparative results for selected cases with/out an extended current-time specification.

The control system for the Poloidal Field (PF) on ITER is a synchronously networked control system, which has several kinds of computational controllers. The Local Control Cubicles (LCC) play a critical role in the networked control system for they are the interface to all input and output signals. Thus, some additional work must be done to guarantee the LCCs proper operation under influence of faults. This paper mainly analyzes the system demands of the LCCs and faults which have been encountered recently. In order to handle these faults, decoupled service-based software architecture has been proposed. Based on this architecture, fault detection and system recovery methods, such as redundancy and rejuvenation, have been incorporated to achieve a fault-tolerant private network with the aid of QNX operating system. Unlike the conventional method, this method requires no additional hardware and can be achieved relatively easily. To demonstrate effectiveness the LCCs have been successfully tested during the recent PF Converter Unit performance tests for ITER.

The EAST poloidal field (PF) control system was upgraded in 2015 and the new system has been in use for the 2015 EAST campaign. This paper presents the implementation of a hardware-in-the-loop (HIL) simulation platform of the EAST PF converter system based on the RT-LAB simulation environment, which was used to improve and evaluate the performance of the real controller. The EAST PF power supply system and its operational modes are presented in this paper. The experiments on HIL simulation platform show that the control algorithms and the over current protection of the controller meet the design requirements well. In addition, the effectiveness of the designed control system has been verified by actual application during the EAST campaign at 2015 for six months.

This paper describes the prototype design of an external bypass that provides protection for the ITER poloidal field converter and magnets in fault conditions. The maximum peak pulse current of the bypass will reach to about 350 kA in the worst case. Three kinds of arm structures are compared according to the design requirement and the specific circumstances of branches. The structure, whose current flows in and out on the same side, is selected and combined with the installation and location. The external bypass structure model is analyzed and compared through analytical and numerical methods. Then a 3-D numerical model of an electromagnetic field is established based on Maxwell. Now a full prototype has been manufactured by China, and the dynamic stability test of the prototype has also been accomplished successfully. It shows that the prototype design can fully satisfy the design requirement and the reliability has been improved.

The ITER poloidal field (PF) magnet system is a set of six circular coils attached to the periphery of the toroidal field coil (TFC) structure. While the manufacturing of the PF coils is being launched in three different countries, the ITER Organization (IO) is looking ahead at assembly of PF coils in the tokamak. The PF coils can be grouped in three pairs with respect to their similarities in design and attachment to the TFC, which leads to preparation of different strategies and tooling for the transportation and assembly. The procedures shall also consider the integration, assembly sequence, and environment of the whole tokamak. For example, the two lower coils PF6 and PF5 must be brought in storage position in the tokamak pit and stay there until all the TFCs are assembled, before being attached to them. This paper presents the assembly scenario of the PF coils, including logistics, analysis of assembly tolerances, aligning strategy, and detailing the particularities of each PF coil, from the moment they are delivered to IO until they are connected to the feeder, waiting for the commissioning of the machine.

The Russian Scientific R&D Cable Institute was obliged to produce and to deliver 20% toroidal field (TF) conductors and 20% poloidal field (PF) cables for the ITER magnet system. The task has been completed in October 2015. This work was a great technological challenge in the development of tricky technologies. Development of manufacturing processes was accompanied by extensive scientific studies directed to an improvement of technologies and a search of causes resulting in change of cables and conductors properties during manufacturing and testing. In this paper, we present the extended review and summary of results obtained for Russian superconductors that permitted improving technologies and finding out the reasons of changes of properties of cables and conductors during manufacturing and testing. The study of rotation and untwisting of TF cables during insertions into jackets allowed developing the model describing the rotation process and suggesting the device which keeps untwisting within demanded levels. Analyses of mandatory SULTAN high-field electromagnetic tests of PF and TF conductors from Russia confirmed good quality and adequacy of manufacturing processes. Micrographic studies of Nb 3Sn strands before and after SULTAN tests permitted qualitatively understanding the reason for the stability of RF TF conductor behavior during multiple electromagnetic and thermal cycling. Statistical studies of residual resistance ratio (RRR) of Nb 3Sn and NbTi strands permitted tracking of its degradation during manufacturing processes and SULTAN tests. The optimal heat treatment process for Nb 3Sn conductors has been confirmed after RRR statistical study. The hydraulic performance has been measured in the short samples of TF conductor. By using the parameters obtained, it is possible to predict hydraulic performance of helium flow in a conductor at any temperature. These results led to deeper understanding and to improving manufacturing processes and quality of TF and PF conductors for the ITER magnet system and could be useful for suppliers of superconducting cables and conductors using ITER-like technologies.

The ITER magnet system is made up of four main subsystems: the 18 toroidal field (TF) coils, the central solenoid, the six poloidal field (PF) coils, and the correction coils (CCs). The feeder system, with its main busbar (MB) and CC busbar (CB), represents one of the main magnet components as well. All coils and busbars with different dimensions used cable-in-conduit conductors. China has signed three conductor packages, which are the so-called procurement arrangements, between ITER and the Chinese Domestic Agency (CN DA): a TF conductor package, a PF conductor package, and a CC and feeder conductor package. They include 7.5% of the TF conductors (11); all the PF2 (12), PF3 (16), PF4 (16), and PF5 (16) conductors; all the CC (18) conductors; and the MB (3) and CB (2) conductors for the feeders. Complex technologies have been developed by ASIPP for the serial production of all ITER conductors, in terms of cabling parameter design, welding, and elaboration of cable insertion, compaction, and winding processes. China has finished all qualification phases and is well into the main series production. All conductor samples required for quality control have successfully passed the SULTAN tests with good performance. The status of the production of ITER conductors in China is described in this paper.

This paper introduces the architecture and characteristics of the poloidal field power supply, and describes the architecture of its control system, which is composed of data acquisition management system, operation monitoring system and real-time control system. According to the characteristics of the control system, a data acquisition management system based on Qt, an operation monitoring system based on EPICS and the real-time control system is based on Linux real-time kernel are designed. The control system realizes real-time control of EAST poloidal field power supply system in 1 ms cycle time, and the equipment status detection and monitoring, data acquisition, storage and inspection. Several experiments have proved that the control system definitely meets the requirements of the poloidal field power supply control system in the complex situations under magnetic fields.

This paper mainly introduces the research and development (R&D) of the high-power dc reactor prototype, whose functions are to limit the circulating current and ripple current in the ITER poloidal field (PF) converter. It needs to operate at rated large direct current 27.5 kA and withstand peak fault current up to 175 kA. Therefore, in order to meet the special requirements of the dynamic and thermal stability, a new prototype design structure of dry-type air-core water-cooling reactor with epoxy resin casting technique is presented, which is based on the theoretical analysis, finite-element simulation calculation and small prototype test verification. Now the full prototype has been fabricated by China industry, and the dynamic and thermal stability tests of the prototype have also been accomplished successfully. The test results are in compliance with the design and it shows the availability and feasibility of the proposed design, which may be a reference for relevant applications.

The Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP) is experienced in the design and fabrication of high-temperature superconductor (HTS) current leads such as those successfully used in the Experimental Advanced Superconducting (EAST) Tokamak. The aim of the 52-kA trial lead was to develop a prototype for the leads supplying the poloidal field and central solenoid magnets of ITER. This lead differs from former HTS current leads and from another trial (68 kA) lead developed in parallel for the toroidal field magnet, in that it features a foil-type heat exchanger, a helium-cooled transition section, and a binary shunt made of stainless steel and phosphor bronze. The 52-kA trial lead was fabricated and tested in the EAST Tokamak facility. Results of the test are presented. An analysis of these results indicates that the design concepts introduced are validated. The test also served to validate the HTS stacks that equipped the lead. These were supplied by InnoST.

The Sun occasionally goes through Maunder-like extended grand minima when its magnetic activity drops considerably from the normal activity level for several decades. Many possible theories have been proposed to explain the origin of these minima. However, how the Sun managed to recover from such inactive phases every time is even more enigmatic. The Babcock--Leighton type dynamos, which are successful in explaining many features of the solar cycle remarkably well, are not expected to operate during grand minima due to the lack of a sufficient number of sunspots. In this Letter, we explore the question of how the Sun could recover from grand minima through the Babcock--Leighton dynamo. In our three-dimensional dynamo model, grand minima are produced spontaneously as a result of random variations in the tilt angle of emerging active regions. We find that the Babcock-Leighton process can still operate during grand minima with only a minimal number of sunspots and that the model can emerge from such phases without the need for an additional generation mechanism for the poloidal field. The essential ingredient in our model is a downward magnetic pumping which inhibits the diffusion of the magnetic flux across the solar surface., Comment: Published in ApJ; 6 pages including 4 figures

HL-2M (Li, 2013 [1] ) is a tokamak device that is under construction. Based on the magnetic coils design of HL-2M, four kinds of divertor configurations are calculated by CORSICA code (Pearlstein et al., 2001 [2] ) with the same main plasma parameters, which are standard divertor, exact snowflake divertor, snowflake-plus divertor and snowflake-minus divertor configurations. The potential properties of these divertors are analyzed and presented in this paper: low poloidal field area around X-point, connection length from outside mid-plane to the primary X-point, target plate design and magnetic field shear. The results show that the snowflake configurations not only can reduce the heat load at divertor target plates, but also may improve the magneto-hydrodynamic stability by stronger magnetic shear at the edge. A new divertor configuration, named “tripod divertor”, is designed by adjusting the positions of the two X-points according to plasma parameters and magnetic coils current of HL-2M.

The remote maintenance schemes in a DEMO reactor are categorized by insertion direction, blanket segmentation, and divertor maintenance scheme, and are quantitatively evaluated by analysing the plasma equilibrium. The positions of the poloidal field (PF) coil are limited by the size of the toroidal field (TF) coil and the maintenance port layout of each remote maintenance scheme. Because the PF coils are located near the larger TF coil and far from the plasma surface, the horizontal sector transport maintenance scheme requires the largest part of total PF coil current, 25% larger than that required for separated sector transport using vertical maintenance ports with segmented divertor maintenance (SDM). In the unsegmented divertor maintenance (UDM) scheme, the total magnetic stored energy in the PF coils at plasma equilibrium is about 30% larger than that stored in the SDM scheme, but the time required for removal and installation of all the divertor cassettes in the UDM scheme is roughly a third of that required in the SDM scheme because the number of divertor cassettes in the UDM scheme is a third of that in the SDM scheme. From the viewpoint of simple maintenance operations, the merit of the UDM scheme has more merit than the SDM scheme.