Institute Of Plasma Physics Research Articles

Impact of local plastic deformation on thermo-magnetic instability of high-J c Nb3Sn strand

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.

Development of a cryogenic DC transformer using for the measurement of large current

Superconducting transformer is a typical facility used for inducing ultra-high current for large-scale scientific measurement systems, such as SULTAN-facility at the Centre de Recherches en Physique des Plasma (CRPP) in Switzerland, FRESCA-facility at CERN in Switzerland etc. A new program has been launched at the Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP) to build a new conductor test facility to investigate the transverse pressure dependence of critical current of various superconducting cables using a compact superconducting transformer and cryogenic electromagnetic press systems. The superconducting transformer can induce a secondary current up to 60 kA, at a background magnetic field up to 15 T at 4.2 K. To measure the induced secondary current at cryogenic temperature precisely, three types of NbTi superconducting coils are designed based on the zero-flux principle to characterize the variation of the measured current. In this paper, the design principle of the cryogenic DC transformer and its commissioning results in a preliminary prototype test device are presented. The measurement results show achievements on accuracy of ±0.6% or less and nonlinearity of better than ±0.25%, which demonstrates the feasibility of the cryogenic DC transformer and provides a reliable technology for the secondary current measurement of a 100 kA superconducting transformer used in the program of Super-X test facility at ASIPP.

Design and analyses on the recuperator of the high temperature PbLi experimental platform for CFETR

As a candidate blanket for China Fusion Engineering and Test Reactor (CFETR), the supercritical carbon dioxide (sCO2) cOoled Lithium-Lead (COOL) blanket is under development. To solve the key scientific issue, i.e. MagnetoHydroDynamics (MHD) effect, the high temperature PbLi experimental platform is planned for construction at the Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP). As one of the key components in this platform, the recuperator acts as waste heat recovery and saving electricity power. In this paper, two types of recuperator are designed and analyzed, including the double-tube and hairpin. Firstly, the theoretical and empirical models are used to obtain different designs by changing the structure dimensions and arrangement, then the non-dimensional evaluation criteria is improved and adopted to determine the optimal design that has better comprehensive performance, i.e. higher heat transfer capacity, lower flow resistance to reduce pressure drop and smaller consume volume of liquid metal to increase economy. Finally, the thermal-hydraulic and thermal-mechanical analyses based on the optimal design of the 3D model are further carried out, which obtains the fields of temperature, pressure, velocity as well as stress, and the results can satisfy with the multiple requirements, i.e. outlet temperature under different operating conditions, flow distribution as well as structural strength. The design methodology and analysis results will provide useful technical and data support for the following engineering design and construction of the PbLi experimental platform.

Thermal-hydraulic simulation of the CFETR TF coil operation during a plasma scenario: Gandalf-Fluent and 4C code-to-code comparison

Within the framework of the comprehensive research facility for fusion technology (CRAFT) activities, a prototype toroidal field (TF) coil has been designed and will be manufactured at the Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP), in Hefei. The main purpose is to develop and validate the manufacturing technologies of the large-scale superconducting coils of the China Fusion Engineering Test Reactor (CFETR). To ensure the safe operation of the CFETR device, the Gandalf-Fluent tool and the 4C code are employed to evaluate the performance of the superconducting magnets, and their thermal-hydraulic simulation results are compared with each other. The comparison shows that the thermal coupling between turns and pancakes and the transient model of the thermal coupling to the casing structures, accounted for in 4C but neglected in Gandalf-Fluent, have a significant effect on the temperature profile. Then a detailed reassessment of the minimum temperature margin (ΔTmarmin) is carried out using the 4C code. The results show that the ΔTmarmin is ∼ 1.0 K (if a realistic value for the inter-turn and inter-pancake coupling parameter is assumed), lower than the design requirement of 1.2 K, and it deserves the attention of the designers.

Performance test of REBCO CICC sub-cables with 10 kA current under 20 T background field

While commercially manufactured rare earth barium copper oxide (REBCO) tapes show significant promise in facilitating the operation of fusion magnets with magnetic fields above 15 T, the design and development of highly stable cable in conduit conductor (CICC) technology is very important to achieve their practical application. To find a good solution for this demand, the Institute of Plasma Physics, Chinese Academy of Sciences, proposed two kinds of CICC design concepts, which are both manufactured from a sub-cable formed by winding REBCO tape around a stainless steel spiral tube. As part of the ongoing activities to develop an REBCO CICC, two sections of sub-cable specimens were manufactured and bent into a U-shape for testing under magnetic fields up to 20 T. A sub-cable specimen with 30 commercial 4 mm wide REBCO tapes displayed around 10 kA at 4.2 K and a background magnetic field of up to 20 T. It also showed stable operation under an electromagnetic (EM) load of around 200 kN m−1, which is above the 150 kN m−1 required by the designed CICC sub-cable. However, the calculated I c of the other specimen degraded from 8.8 kA to 8.5 kA when cycling with an EM load of around 160 kN m−1. The lower calculated n-value at 77 K and self-field as well as the observed imprints on the disassembled tape edges suggested that defects were generated in the cable during cabling, bending to the sample holder or operation with high EM and thermal loads. These results exhibit the potential and feasibility of using high flexible REBCO cable (HFRC) sub-cables for high-field fusion magnets. However, the winding parameters need to be optimized to ensure safe operation in more complex conditions, such as in tokamaks, especially if using tapes similar to those used in sample-B in this study. Moreover, it is imperative to establish much more rigorous requirements for coil manufacturing processes in order to avoid the occurrence of defects in the tapes.

Design, manufacturing, evaluation, and performance of a 3D-printed, custom-made nozzle for laser wakefield acceleration experiments.

Laser WakeField Acceleration (LWFA) is extensively used as a high-energy electron source, with electrons achieving energies up to the GeV level. The produced electron beam characteristics depend strongly on the gas density profile. When the gaseous target is a gas jet, the gas density profile is affected by parameters, such as the nozzle geometry, the gas used, and the backing pressure applied to the gas valve. An electron source based on the LWFA mechanism has recently been developed at the Institute of Plasma Physics and Lasers. To improve controllability over the electron source, we developed a set of 3D-printed nozzles suitable for creating different gas density profiles according to the experimental necessities. Here, we present a study of the design, manufacturing, evaluation, and performance of a 3D-printed nozzle intended for LWFA experiments.

High-Performance FPGA Streaming Data Concentrator for GEM Electronic Measurement System for WEST Tokamak

The paper presents developments and significant improvements of the soft X-ray measurement system installed at the WEST tokamak. In the introduction, a brief discussion is carried out in the scope of energy shortage, fusion energy as a remedy, and the necessity of impurities monitoring in the scope of stable and long plasma discharge. This requires high-speed and accurate measurement systems due to the intense data streams that need to be processed online. For that reason, the SXR GEM FPGA-based system was designed by the Institute of Electronic Systems of the Warsaw University of Technology, the Institute of Plasma Physics and Laser Microfusion from Warsaw, and installed at the WEST tokamak in collaboration with the Institute for Magnetic Fusion Research, Commissariat à l’Énergie Atomique. It is the second-generation system, whereas the first was installed at JET tokamak. The article describes the architecture of the designed electronic part of the system. The paper presents an entirely new approach for the data concentration module implemented in FPGA. The main premise is to select only active data and send them chronologically to the embedded computer with high throughput. It is the essential component for long-term plasma operations, about 1 min, now carried out at WEST tokamak during the C3 campaign. The paper describes the laboratory tests under the exploitation of various radiation sources and the implementation of the solution and measurements during tokamak plasmas. Previous and current data acquisition methods are compared. The results show that implementing a new local trigger has significantly improved the system performance compared to the global trigger-based acquisition. The results are approximately 17 times better in the scope of performance and more than 20 times better in terms of data compression. The described design was successfully applied during the most recent 2023 experimental campaign at the WEST tokamak.

Nonexcessive-ΔV and Low Complexity Model Predictive Control Based on Finite-State Machine for Three-Level Three-Phase Inverters

For three-level three-phase inverters, a novel finite-state machine based model predictive control (FSM-MPC) is proposed. The state transition diagram of FSM is regarded as the operation guideline to avoid excessive voltage jumps (Δ <italic xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">V</i> ). No more than five voltage vectors (VVs) are selected as FSM-MPC's candidates VVs, based on the VV reference and the previous optimal VV. And the cost function is simplified in terms of time duration. Thus, evaluating a candidate VV requires a very low computation cost. Without losing control performance, the proposed FSM-MPC's execution time is decreased to 50% of the MPC algorithm, which enumerates all basic VVs to get global optimal VV and performance. Compared to the existing simplified algorithms, the proposed FSM-MPC makes current harmonics lower, and its average switching frequency is 33% less, which means much less switching loss. Furthermore, the proposed algorithm is robust when the electric-circuit parameters are mismatched in the control system. Experimental results are provided to validate the advantages of the proposed algorithm. <fn fn-type="other" id="fn3" xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"> This work was supported in part by National Natural Science Foundation of China under Grant 52177204, and in part by Natural Science Foundation of Hunan Province under Grant 2021JJ30875. (Corresponding author: Jian Yang) H. Zhou, J. Yang, D. Song and M. Dong are with the school of Automation, Central South University, Hunan 410083, People's Re-public of China (e-mail: ; ; ; ). L. Huang and X. Chen are in Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031, People's Republic of China (e-mail: ; ). </fn>

Development of a Scalable MMC Pulsed Power Supply through HIL Methodology

Nuclear fusion experiments are becoming one of the most interesting focuses of research, given the hope of generating programmable, safe, and green energy. Among them, ASDEX (axially symmetric divertor experiment) upgrade has been operating at the Max Planck Institute for Plasma Physics (IPP) research center since 1991. To ignite and confine the plasma, several coils must be supplied through controllable high-current pulsed power supplies. The toroidal field magnets are here considered and a modular multilevel converter (MMC)-like system was designed and tested thanks to a small-scale prototype in previous works. The MMC-like topology, consisting of full-bridge submodules (SMs) with IGBTs and supercapacitor and exploitable also for other industrial applications, was chosen because of its modularity, redundancy, fault tolerance, and large amount of stored energy. The prototype, made of four SMs, was necessary to highlight critical key points in the design process. However, its scalability must be further tested before building a full-scale power supply, meant to reach almost 2400 SMs to guarantee the energy required by the load. This paper aims at validating hardware-in-the-loop (a powerful, safe, and relatively inexpensive real-time simulation environment that enables testing with real control boards) as a useful technology for power supply scalability studies and not only for control strategy tests. The results obtained previously from the prototype will allow us to finally increase the number of SMs and test the MMC-like scalability.

COMPASS-U Global Heat Balance Calculations

COMPASS-U is a medium size, high magnetic field experimental tokamak ( <inline-formula xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"> <tex-math notation="LaTeX">$R =0.9$ </tex-math></inline-formula> m, <inline-formula xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"> <tex-math notation="LaTeX">$B_{t} = 5$ </tex-math></inline-formula> T, and <inline-formula xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"> <tex-math notation="LaTeX">$I_{p} = 2$ </tex-math></inline-formula> MA), built at the Institute of Plasma Physics, Czech Academy of Sciences (IPP-CAS). This global heat balance calculation for COMPASS-U was done at Princeton Plasma Physics Laboratory (PPPL). Based on our previous experience of building global thermal model for National Spherical Torus Experiment-Upgrade (NSTX-U) at PPPL, this 2-D global thermal model geometry represents a typical cross section of COMPASS-U machine. The model includes thermal radiation, conduction, and convection among components and also between the machine and outer environment. Helium gas heating/cooling was modeled with fluid element and surface element. This model was used to calculate component temperatures and heat distribution during heat up, cryogenic cool down, normal operation, and fault operation scenarios. Definition of 14 main thermal scenarios was provided by IPP-CAS. First nine are normal operation scenarios. Scenarios #10–#14 are fault scenarios. Some results of thermal scenario #5 will be given and discussed in this article, including peak temperatures, temperature ratcheting, energy distribution, and required cooling power.

Capabilities of Thomson parabola spectrometer in various laser-plasma- and laser-fusion-related experiments

Abstract The Thomson parabola spectrometer (TPS) [1] is a well-known, universal diagnostic tool that is widely used in laser plasma experiments to measure the parameters of accelerated ions. In contrast to other popular ion diagnostics, such as semiconductor detectors or ion collectors, the TPS is not greatly affected by electromagnetic pulses generated during high-power laser interaction with matter and can be tuned to acquire data in various energy ranges of accelerated ions, depending on the goal of the experiment. Despite the many advantages of this diagnostic device, processing the collected data is a difficult task and requires a lot of caution during interpretation of gathered results. In this work, we introduce the basic principles of operation and data analysis based on the numerical tool created specifically for the TPS designed at the Institute of Plasma Physics and Laser Microfusion, present a range of data obtained during various recent experiments in which our TPS was used, and highlight the difficulties in data analysis depending on the purpose of the experiment and the experimental setup.

Introduction of an inventory tool for the utilities and diagnostics of ASDEX Upgrade

Until now, more than 40,000 plasma discharges have been carried out at the Tokamak experiment ASDEX Upgrade of the Max Planck Institute for Plasma Physics (IPP) at Garching / Germany. In order to achieve the scientific objectives, the power supply and additional heating systems as well as the plasma diagnostics have been continuously expanded and a huge variety and quantity of facilities is available in the torus hall. Over time, it becomes increasingly difficult to keep track of all components, their physical location and to identify relationships and dependencies between individual plants. In particular, the mutual interference and a correct grounding of the installation are of essential interest in order to guarantee good signal quality of the diagnostics. On the other hand, current regulations for plant safety and personal protection must be complied with.In order to meet these requirements and to obtain a better overview, a new inventory tool has been introduced. Essential information about the current conditions, parameters, logical dependencies, maintenance intervals, contact persons, overview diagrams, physical location and plant pictures can be stored. Successively, all these data will be accessible at a single place.The paper describes the adaptation of a database system designed for IT applications towards an inventory tool for a matured research facility. It presents the scope of services, first experiences of the users applying the new tool and future developments.

Development of High Current Density, Compactness NbTi Superconducting Coil for the Maglev System

A 176 A/mm <sup xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">2</sup> , compactness racetrack NbTi superconducting coil has been developed in Institute of Plasma Physics, Chinese Academy of Sciences. The purpose of this research is to further enhance the levitate and guide force for superconducting maglev. The coil was wound monolith rectangle NbTi/Cu superconducting wire, insulated by fiberglass and polyimide, and cured by epoxy with vacuum pressure impregnation technology. This article pays attention to the coil design, fabrication, and “training” test effect associated with the use of epoxy impregnates. The coil performed the quench training fourteen times and the current increase to 330A@5.9T under 4.6 K. The test process will be outlined in detail, the “training” results will be discussed in this article.

Structure, design, and test of 13 kA HTS current lead

The 13 kA High Temperature Superconducting Current Leads (HTSCL) were designed and manufactured by the Institute of plasma physics, Chinese Academy of Sciences (ASIPP). This lead differs from former HTS current leads, in that it features an integrated heat exchanger (HEX) structure, a HTS magnetic shield and protective sleeve, and a Low Temperature Superconducting (LTS) module with a through cold conduction channel. The HTS and LTS module is cooled by SHe. In November 2020, the 13 kA steady-state test, Loss Of Flow Accident (LOFA) test, joint resistance test, pulse current test, insulation voltage test and other low-temperature electrical test items of the leads were completed. The test results indicate that all parameters meet the design requirements.

Combined Normal and Disruption, Electromagnetic Transient, Thermal, and Structural Analysis of COMPASS Upgrade

COMPASS Upgrade is a replacement for the COMPASS device at the Institute of Plasma Physics (IPP) of the Czech Academy of Sciences. It is a cryogenic copper machine with a major radius of 0.89 m, with 5 T at the plasma centerline and 2-MA plasma current. It is intended to develop ITER relevant plasma densities and high power fluxes in the divertor region. The entire vacuum vessel is planned to reach 500°C. The COMPASS-U TF coil is a bitter plate-like design with radial conductor plates that extend the width of the TF inner leg. Currents redistribute radially based on resistive and inductive effects. This necessitates an electromagnetic (EM) transient analysis of the coil current distribution. For copper coils like COMPASS Upgrade, the solution must be coupled with a thermal solution to properly include the temperature-dependent resistive effects. The EM transient solution includes the TF magnetic loads, and a stress pass on the results produces in-plane loads. The EM solution chosen requires inclusion of all the conducting and nonconducting regions surrounding the TF coil. With the TF current distribution solved, the addition of the poloidal coils and resulting background fields allows the determination of the out-of-plane loads. Integration of the Joule heating produced from the EM solution produces temperature distributions throughout the shot, which can be read in to a structural pass to include thermal stresses in the evaluation. This is true of both TF and PF temperature prediction with appropriate packing fractions. Currents in the PF coils, with computed background fields, produce Lorentz loads. Time transients of the PF currents produce startup eddy currents in the structures included in the model; inclusion of the vessel and other passive structures allows these to be included in a stress pass. The EM transient solution includes all the elements needed for a disruption analysis with prescribed motions and quench of the plasma. From a model originally intended to simulate current diffusion in the TF coil, the analysis can be refined to be used to predict the performance of most of the components of the tokamak or provide boundary conditions for more detailed submodels. In this article, the use of the current diffusion model for bounding assessments of the COMPASS Upgrade TF, PF, vessel, external structure, and disruption analyses will be presented.

Hydraulic characteristic analysis of a fully internal-flow cooling cryosorption-panel for CFETR torus cryopump

The China Fusion Engineering Test Reactor (CFETR) will develop a high-performance cryopump for pumping the torus vaccum vessel. Cryosorption-panel is the key component of torus cryopump, which is applied to adsorb gaseous molecules in cryogenic conditions. Considering the special geometries of cryosorption-panel, a research campaign has been carried out by Institute of plasma physics, Chinese Academy of Sciences (ASIPP) to analyze the structure and assess the hydraulic performance of the cryosorption-panel. This research activity was performed by Finite Element Method (FEM) and Computational Fluid Dynamics Method (CFD). In this article, the cryosorption-panel structure was designed and analyzed, and five representative fluid fields with different geometries were chosen, modeled and numerical simulated. The inlet-outlet pressure drops and flow distribution data under mass flow rates between 0.03kg/s and 0.2kg/s was collected and quantitatively analyzed. By analyzing the simulation results, the influence of fluid field thickness and pin fin arrangement on hydraulic performance was summarized. For the cases with fluid field thickness of 4mm, 6mm, 8mm, the functional relationship of hydraulic resistance coefficient ξ versus Reynolds number (Re) was calculated respectively: ξ = 75.26Re−0.07475; ξ = 120.7Re−0.1464; ξ = 385.3Re−0.2876;

Mechanical effects: challenges for high-field superconducting magnets.

Due to its clean products and sufficient raw materials, fusion energy is expected to become one of the main solutions of the energy crisis and ensuring the sustainable development of human society, which is a long-term strategic frontier field. The promise of fusion energy is to constrain the motion of high-temperature plasma by the high magnetic field generated by superconducting magnets, and then achieve controllable thermonuclear fusion. Fusion power is proportional to the fourth power of the magnetic field strength. Thus, future commercial fusion reactors need a higher magnetic field as the basis for sustainable development [1]. In order to verify the scientific and technological feasibility of fusion energy, China, the United States, the European Union, Russia et al. have jointly participated in the construction of the International Thermonuclear Fusion Test Reactor (ITER), which is expected to produce the first plasma discharge by 2025 [2]. Currently, China is leading the world in many fields of fusion energy research. For example, the experimental advanced superconducting Tokamak (EAST) whole-superconducting Tokamak located at the Institute of Plasma Physics in the Chinese Academy of Sciences has achieved a repeatable world record of stable plasma operation at 120 million degrees Celsius for 101 seconds, which provides a solid foundation for ITER and also China's future Independent Building Fusion Reactor (https://www.cas.cn/syky/202105/t20210528_4790357.shtml). Prof. Jiangang Li, an academician of the Chinese Academy of Engineering, participated in and completed the design and construction of EAST plasma facing componments (PFCs) engineering by the support of the national '9th five-year plan' major scientific and technological infrastructure, and presided over the completion of the national '11th five-year plan' major scientific and technological infrastructure-EAST auxiliary heating system project. He also hosted the national '13th five-year plan' major scientific and technological infrastructure-Integrated Research Facility for Critical Systems of fusion reactor comprehensive research facility for fusion technology (CRAFT). Many important scientific and technological problems have been solved and overcome by Prof. Li and his co-workers, which puts China's plasma physics research and fusion engineering technology at the forefront of global engineering.

Performance of first insert coil with REBCO CICC sub-size cable exceeding 6 kA at 21 T magnetic field

The Institute of Plasma Physics at the Chinese Academy of Sciences is developing the REBCO cable in conduit conductor (CICC) technology for applications in next-generation nuclear fusion devices. The aim is to develop a CICC comprised of six REBCO sub-cables to satisfy the requirements of operation with a current of around 40 kA and a peak field of up to 20 T. To qualify the performance of the sub-size REBCO cable to be used in the CICC, two 25-turn insert solenoids have been designed, manufactured and tested at a current exceeding 6 kA subjected in a background field supplied by a water-cooled resistive magnet. The insert solenoid, wound from a 11.5 m long REBCO CORC® cable, was designed to investigate its current carrying capacity under high field and electromagnetic (EM) load at 4.2 K. Tests were performed under a background magnetic field up to 18.5 T, resulting in a peak magnetic field on the innermost layer turns of around 21.1 T at an operating current of 6.3 kA. The effects of operation with cyclic EM loads were tested by repeated current ramps to around 95% of the critical current. Moreover, the V–I characteristics were measured at 77 K and the self-field, to check the effects from warm-up and cool-down (WUCD) cycles between room temperature and 77 K with liquid nitrogen. The results show no obvious degradation after dozens of high-current test cycles in background fields ranging from 10 T to 18.5 T. The insert solenoid demonstrates the stable operation of the REBCO sub-size cable for CICC with EM loads of about 90 kN m−1 and WUCD cycles between room temperature and 77 K. These promising results indicate the potential of this technology for further applicationsin particular, for full-size CICC for high-performance fusion magnets.

Feasibility Study of ITER In-Vessel Coils Bracket Manufacture and Integration

Feasibility study of ITER In-vessel coils bracket manufacture and integration had been developed in Institute of Plasma Physics, Chinese Academy of Sciences. The ITER In-Vessel Coil system is comprised of Edge-Localized Mode (ELM) and Vertical Stabilization (VS) coils. The ELM coils are used to mitigate the Edge Localized Modes and the VS coils are used to provide Vertical Stabilization of the plasma. Designed bracket for IVC coils is a kind of building block type three or four stacked components with arcuate groove matching with round conductor. This paper describes structure design, manufacture and integration process of the ELM and VS bracket. R&D of bracket weld and assembly sequence optimization are carried out to determine the welding and assembly process. At last, three brackets are integrated in ASIPP.

Development of a 7 T NbTi solenoidal magnet with 72 mm aperture

A close-packed coil with a 72 mm aperture is developed using the formvar-insulated NbTi wire made from the WST company. The micrometer-level inter-turn spaces are filled with CTD-101 K utilizing a vacuum pressure impregnation method with a controllable flow speed of the epoxy resin. The cryogenic excitation results indicate that the coil inter-turn spaces are well filled, so that it can generate a 7.35 T central magnetic field at an operation current of 176 A with no coil quench. The maximum loadline of the coil is proved to be 98.96%. The maximum hoop-stress is approximately 64.1 MPa at the center of the inner most coil layer. The VPI method is applicable to the close-packed coils (tcoil ≤ 28 mm) wound from the formvar-insulated NbTi wire.Differ from the fiberglass-insulated Nb3Sn wire, the formvar insulated NbTi provide a relative smooth wire surface, which does not absorb the epoxy resin. This significantly increases the difficulties of vacuum pressure impregnation (VPI) for a thick close-packed NbTi coil. With the purpose of understanding the applicable of current VPI procedure on formvar insulated NbTi coils, a close-packed solenoidal magnet wound from the formvar insulated NbTi is wound, vacuum pressure impregnated and cryogenic measured at the Institute of Plasma Physics (ASIPP), Chinese Academy of Sciences (CAS). The design aperture of the magnet is 72 mm, with coil thickness of 28.75 mm and coil height of 180.5 mm, the magnet is expected to generate a 7.04 T central magnetic field with an operation current of 170 A at 4.2 K. The excitation results show that the central magnetic field of the NbTi coil impregnated with current VPI procedure is able to reach 7.35 T with no quench under an operation current of 176 A. The loadline is close to 98.96% of the NbTi wire.