In the framework of the program to fabricate a high-field superconducting magnet of the order of 19 T @4.2 K with a 150 mm free bore, a 15 T hybrid solenoidal superconducting magnet with a 78 mm bore was developed and preliminary tested. The purpose of the program is to upgrade the material test facilities in the Institute of Plasma Physics Chinese Academy of Science (ASIPP) and further extend their measurement functionality. This paper presents the design of the 15 T hybrid magnet and its preliminary test results. The magnet comprises two high-performance Nb3Sn coils, one ITER Nb3Sn coil and one NbTi coil respectively for the high-, medium- and low field section. The axial length of the homogeneity within a 99 % magnetic field region is designed to more than 40 mm, and the ratio between the peak and central magnetic field is expected to be less than 1.025. The Nb3Sn coils were reacted separately, after which all coils including the NbTi coil were impregnated with CTD-101 K individually. Then, the impregnated sub-coils were assembled into a concentric-coaxial configuration and electrically connected in series. The hybrid magnet was inserted to a liquid helium bath for the excitation test, it finally reaches 97.8 % of the design current with a 14.84 T central magnetic field.

The Central Solenoid Model Coil (CSMC) project was launched in the Institute of Plasma Physics Chinese Academy of Sciences for the China Fusion Engineering Test Reactor (CFETR). The CFETR CSMC is designed to generate 12 T peak magnetic field when the running current is 47.65 kA. The CSMC is composed of two Nb3Sn coils in the internal high field and three NbTi coils in the external low field. Electrical joints are one of the key technologies for the CSMC, because the superconducting coils are electrically connected into a single circuit by the joints. Lap type joints have been designed for the CSMC for its proven reliability and easy assembly. It is a great challenge to develop special equipment and technologies for the joint manufacture, including vacuum brazed terminal sleeve assembly, jacket removal, chrome coating removal for Nb3Sn cable, nickel coating removal and silver plating for NbTi cable, compact termination, etc. This paper mainly focuses on the research and development activities of the joint manufacture and the solution of the technical issues. The techniques and equipment can provide a basis for the electrical joint manufacture for the CSMC and other large-scale superconducting magnets.

大孔径高均匀度高场磁体是进行超导材料性能测试的必要条件之一.本文主要介绍了15 T高场复合超导磁体的低温励磁实验与结果.该磁体包括两个高性能Nb3Sn线圈、一个ITER Nb3Sn线圈和一个NbTi线圈,分别处于磁体的高、中、低场部分.磁体外径329.9 mm,高度401.44 mm,中心孔径77.5 mm,磁体中心磁场测量值14.94 T,轴向磁场(98％)均匀区测量长度55 mm,超过了磁体设计指标要求的40 mm,测试结果满足设计要求.

The Comprehensive Research Facility for Fusion Technology (CRAFT) project has been launched in 2019, for developing the essential engineering technologies for Chinese Fusion Engineering Testing Reactor (CFETR). Within this project, a full-size toroidal field (TF) coil will be built as the prototype coil for CFETR. Based on design of CFETR magnet system, the TF coil will operate at 95.6 kA in a peak field of 14.5 T. The high-J(c) Nb3Sn strand is taken into consideration due to the critical current density of ITER-grade Nb3Sn is too low at 14.5 T. Considering that it will be the first time to apply the high-J(c) Nb3Sn strand in the large-scale cable-in-conduit conductor (CICC) for fusion magnet, a conductor sample made of high-J(c) Nb3Sn strand with short twist pitch (STP) cable pattern was manufactured in ASIPP and tested in SULTAN facility, to investigate the feasibility. The test campaign focuses on the impact of cyclic electromagnetic (EM) loading and warm-up cool-down (WUCD) to the performance of the conductor, the strain distribution of the conductor before and after EM cycles was measured by inductive method to make a deeper insight of the conductor performance evolution. AC losses tests have also been carried out, providing relevant information for further coil design.

China fusion engineering test reactor (CFETR), based on ITER technology and bridged between ITER and DEMO, has been supported by China government to start technologies R&D and engineering design. The field of CFETR at plasma core is 6.5 T, maximum field of TF coil is about 14.8 T. TF coil is wound by Nb3Sn and NbTi CIC conductors. Coil weight is about 650 tons with height about 21.7 m and width about 12.3 m. Prototype TF coil manufacture design consists of mechanical & electro-magnetic design and analysis, conductor design and analysis, coil AC loss analysis, thermal–hydraulic analysis and coil cooling, quench detection and coil protection, coil winding, case manufacture, and coil assembling. The TF prototype coil is one sub-task of the CRAFT project, which will last for 5 year and 8 months and is constructed by ASIPP and SWIP cooperatively. The preliminary design of the TF prototype coil has been finished at the present stage. In this paper, the basic design requirements of TF magnet are introduced firstly. Next, the preliminary design and analysis of the TF coil is described, which includes the winding package (WP) design, the coil case (CC), the conductor design and analysis, the electromagnetic, mechanical and AC losses analysis. Then the process and tooling design for the WP and CC manufacturing and assembling is presented. Finally, a summary and a plan are given.

Since the last IAEA Fusion Energy Conference in 2016, the EAST physics experiments have been developed further in support of high-performance steady-state operation for ITER and CFETR. First demonstration of a >100s time scale long-pulse steady-state scenario with a good plasma performance (H-98(y2) similar to 1.1) and a good control of impurity and heat exhaust with the upper tungsten divertor has been achieved on EAST using the pure radio frequency (RF) power heating and current drive. The EAST operational domain has been significantly extended towards a more ITER and CFETR related high beta steady-state regime (beta(p) similar to 2.5 and beta(N) similar to 1.9 of using RF and NB and beta(p) similar to 1.9 and beta(N) similar to 1.5 of using pure RF). A large bootstrap current fraction up to 47% has been achieved with with q(95) similar to 6.0-7.0. The interaction effect between the electron cyclotron resonant heating and two lower hybrid wave systems has been investigated systematically, and applied for the improvement of current drive efficiency and plasma confinement quality in the steady-state scenario development on EAST. Full edgelocalized mode (ELM) suppression using the n = 2 resonant magnetic perturbations has been achieved in ITER-like standard type-I ELMy H-mode plasmas with a range of the edge safety factor of q(95) approximate to 3.2-3.7 on EAST. Reduction of the peak heat flux on the divertor was demonstrated using the active radiation feedback control. An increase in the total heating power and improvement of the plasma confinement are expected using a OD model prediction for a higher bootstrap fraction. Towards a long-pulse, high bootstrap current fraction operation, a new lower ITER-like tungsten divertor with active water-cooling will be installed, together with further increase and improvement of heating and current drive capability.

Modified stainless steel 316LN is selected as a candidate material for the China Fusion Engineering Test Reactor (CFETR) central solenoid model coil (CSMC) because of the high strength combined with good ductility at cryogenic temperature. The tensile properties, fatigue crack growth rate, and fracture toughness of the SS316LN tube in solution-annealed and aged (575 degrees C/100 h and 650 degrees C/100 h) conditions were evaluated at 4.2 K. The fatigue crack growth and tensile properties for the solution annealed conduit were high enough to satisfy the design requirements for CFETR CSMC. However, the fracture toughness of the aged conduit is not satisfied, since there was a significant decline from 280 to 110 MPa-m(1/2) after cold working and aging treatments. The chemical compositions and fractures have been analyzed to assess the reason and recommend modifications that could improve fracture toughness and fatigue crack growth properties.

High manganese steel is a promising choice as a jacket material for Nb3Sn superconducting magnet due to the low thermal expansion, high strength combined with good ductility. One kind of high manganese steel with the Mn content of about 21% was developed for reviewing the mechanical properties at 4.2 K. In this paper, the mechanical performances of the developed high Mn steel in solution annealed as well as in cold worked and aged condition were tested. Influences from cold working and aging processes on microstructure and mechanical properties were analyzed.

The ITER IVCs are made of stainless steel mineral-insulated conductors composed by an OFE copper conductor and a SS316LN jacket. The baseline design requires the coils to endure all ITER machine life time. After a brief review of past and recent literatures, no results are available and of our interest. Moreover, cold working, aging treatments as well as the environment temperature affect the mechanical and fatigue properties. The present paper summarizes data from tensile and uniaxial tension load controlled fatigue tests on both the steel jacket and the copper conductor in the IVC representative operation conditions. Tests are performed on specimens sectioned from the tubes undergone cold working of conductor compaction, coil winding and so on as well as heat treatment at 240 degrees C for 24 h of baking, and finally tested at operation temperature of 120 degrees C. The fatigue tests were conducted at frequency of 15 Hz and the stress ratio of R = 0.1. Results from tensile tests show both strength and ductility decrease as temperature rises from RT to 120 degrees C. Results from fatigue tests are presented as S-N relationships with the maximum stress expressed as a percentage of the strength under static loading. It was found that the fatigue strength limit for OFE copper conductor and SS316LN jacket were close to 120 MPa and 275 MPa respectively. The influence from residual stresses on the static and fatigue performance of 316LN jacket were also investigated It was found that the ductility and fatigue life can be improved by residual stress releasing of heat treatment at 650 degrees C for 100 h. (C) 2017 Elsevier Ltd. All rights reserved.

An Nb3Sn cable-in-conduit (CIC) conductor with short twist pitch structure has been developed at Institute of Plasma Physics Chinese Academy of Sciences, Hefei, China, to meet the central solenoid model coil (CSMC) project of China fusion engineering test reactor (CFETR). 12 T magnetic field and 1.5 T/s field change are the most basic design target. The coupling loss characteristics of Nb3Sn CIC conductor is the key design requirement for CSMS because the deposited energy in magnets from ac losses can decrease the safe margin in the nuclear fusion device. In order to evaluate the coupling loss performance of the Nb3Sn CIC conductor, the ac loss measurement for Nb3Sn CIC conductor sample was carried out at SULTAN facility. This paper presents the specific structure design of the Nb3Sn CIC conductor and ac losses test results first. Then, the coupling losses characteristics are discussed with Multizones PArtial Shielding model and the key fit parameters are obtained from measurement results. Finally, two typical field cases are considered to evaluate the coupling loss energy of Nb3Sn CIC conductor for safe operation of the CFETR CSMC and the optimization of the cryogenic system.

ITER is incorporating two types of In-Vessel Coils (IVCs): ELM Coils to mitigate Edge Localized Modes and VS Coils to provide a reliable Vertical Stabilization of the plasma. Strong coupling with the plasma is required in order that the ELM and VS Coils can meet their performance requirements. Accordingly, the IVCs are mounted on the Vacuum Vessel (VV) inner wall, in close proximity to the plasma, just behind the Blanket Shield Modules (BSM). Due to high radiation environment, mineral insulated copper conductors enclosed in a stainless steel jacket have been selected. The reference design and prototype work provided a good basis for the development of radiation resistant conductor capable of operating within the harsh conditions in ITER vacuum chamber. However, this effort identified shortcomings in achieving satisfactory manufacturing solution, and most significantly, difficulties in brazing the brackets onto the ELM coil conductor. Since this process has not proven successful, alternative designs are under development and prototyping. Prototype manufacturing on the alternative designs has been completed at ICAS, Italy and ASIPP, China. The aim was to eliminate the need for internal coil joints, to prove the principle of longer conductor length manufacturing, and to perform bending and welding trials on two different conductor cross-sections: circular and square. The procurement of the IVCs and their conductors will be done via direct call for tender from the ITER Organization. This paper will give an overview of the alternative design and prototype manufacturing of the ITER In-Vessel coils. (C) 2017 ITER Organization. Published by Elsevier B.V. All rights reserved.

Glass-fiber reinforced plastics is commonly designed as insulation materials for fusion magnet coils and is selected for the turn and ground insulation for the China Fusion Engineering Test Reactor (CFETR) Central Solenoid Model Coil (CSMC). The mechanical properties are being subjected to investigations with respect to the design requirements and operating conditions at present. The preliminary designed insulation system for CFETR CSMC mainly consists of S-glass fiber reinforced tape, interleaved with corona-treated Kapton HN, and vacuum-pressure impregnated in a DGEBF epoxy system. For the composite structure, the tensile and shear performance of the insulation system at room temperature and liquid nitrogen temperature (77 K) were assessed. Meanwhile, effects on the tensile and inter-laminar shear performance from cycling in liquid nitrogen were investigated. The failed specimens were observed to verify the insulation system failure mechanisms.

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.

The conductor for the ITER correction coils (CCs) is manufactured in China and is tested in the SULTAN facility. The third China CC conductor for the qualification SULTAN test sample (CCCN3) was prepared from a 200-m-long CC conductor unit length (Phase II). The fourth and fifth China CC conductor samples for SULTAN testing (CCCN4 and CCCN5) were prepared from the 576-m-long production unit lengths (Phase III). The current sharing temperature T-cs test results show that the sections of CCCN3, CCCN4, and CCCN5 have high T-cs performance. Using the electrical method, the T-cs was 7.13 K for CCCN3, 7.10 K for CCCN4, and 7.09 K for CCCN5 at the ITER reference test conditions, i.e., at B-sultan = 3.7 T and current equal to 10 kA (B-peak = 4.0 T). Even after the 2000 electromagnetic load cycles for CCCN3 conductor and 1000 electromagnetic load cycles for CCCN4 and CCCN5, the T-cs performance of all samples did not changed appreciably. Based on the CC conductor sample test results at 10 kA and at different fields, the T-cs was extrapolated to the peak field B-peak = 4.1 T and current equal to 10 kA. The extrapolated T-cs is 7.07 K for CCCN3, 7.05 K for CCCN4, and 7.03 K for CCCN5. The T-cs performance for all conductors is higher than the ITER acceptance criteria of 7.00 K.

The correction busbar (CB) conductors for the ITER project are being produced in China. The second China CB conductor sample (CBCN2, Phase II of conductor testing) and the third CB conductor sample (CBCN3, Phase III testing) were made of the WST NbTi strands, cabled by Changtong, jacketed at ASIPP, and tested in the SULTAN facility. The current sharing (Tcs) test results show that the conductor sections of CBCN2 and CBCN3 have high Tcs performance. Using the electrical method, the Tcs of both CBCN2 and CBCN3 samples was 7.50 K at 10 kA/2.9 T (maximum operating field) before and after electromagnetic loading. AC loss measurement was performed before any electromagnetic loading of the sample, and the measurements were repeated after 1000 or 2000 cyclic loads. The AC loss measurements were performed at 2-T background field, without transport current. The ac loss of the CBCN2 and CBCN3 conductors slightly differs at ±0.2T ac field and various frequencies. According to the SULTAN test result, Tcs of CBCN2 and CBCN3 conductor samples meet the ITER acceptance criteria of 7.0 K at Bpeak field is 2.9 T and 10 kA.

Four ITER main busbar (MB) conductor samples were tested in the SULTAN test facility (Centre de Recherches en Physique des Plasma, Ecole Polytechnique Fédérale de Lausanne, Switzerland) between 2011 and 2013. The MB conductors are NbTi-based cable-in-conduit conductors (CICCs), and they will become part of the feeder system of the ITER magnets. The measured dc performance of the four MB samples varied significantly, supposedly depending on the design of the bottom terminations. Two out of three samples with a U-bend box, made of a continuous conductor section, exhibit approximately 0.5 K lower current-sharing temperature Tcs than the sample with a solder-filled bottom joint, consisting of two straight conductor sections. We assess the theoretically expected Tcs performance of the MB conductor based on the characterization of individual NbTi strands, on the ITER NbTi scaling law, and on the magnetic field distribution across the cable cross section. The magnetic field in a SULTAN sample consists of three components, namely, the background SULTAN field, the self-field generated by the current in the conductor under test, and the magnetic field generated by the current in the return conductor. Taking into account all the three components, we calculate the average electric field in the cable and determine Tcs as in the experiment, namely, as the temperature at which the electric field reaches the critical value of Ec=0.1 μV/cm. The theoretically assessed Tcs confirms that the MB sample with solder-filled joint behaves as expected.

After a combination of temporary modification, transesterification, cyclization and polycondensation, a series of high-Mn and high-Tg poly(isosorbide terephthalate) copolymers were synthesized.

An ITER Organization (ID) Task Agreement (TA) "Final Design and Prototyping of the ITER In-Vessel Coils (IVC) and Feeders" is almost finished by Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP). ITER IVCs consist of edge-localized mode (ELM) and vertical stabilization (VS) coils. One prototype Mid-ELM coil complete with 19 brackets brazed with the conductors and one prototype 120 degrees section of upper VS coil with structural components brazed to the conductors have been fabricated. Compaction method is developed successfully for the mineral-insulated conductor (MIC) manufacture. Approximate 110 m Inconel 625 jacket MICs for Mid-ELM prototype coil and 80m stainless steel 316L jacket MICs for VS prototype coil were manufactured. Most of the copper tubes used for the MICs fabrication failed the ultrasonic testing (UT), but the jacket tubes have good passing rate. Manufacture processes and inspection for the MICs are presented in this paper. (C) 2015 Elsevier B.V. All rights reserved.

The fourth China TF conductor sample (TFCN4) for Phase II, the left leg of the fifth China TF conductor sample (TFCN5) for Phase III, as well as the right leg of the TFCN5 conductor for Phase IV, were made of the WST strands, cabled by Baosheng, jacketed at ASIPP, and prepared by CRPP following the IO-approved procedure and tested in the SULTAN facility. The T cs test results show that both legs of the TFCN4 and TFCN5 have high T cs performance. Using the electrical method, the T cs was 6.56 K for the left leg of TFCN4 and 6.58 K for the right leg of TFCN4 at 68 kA/10.78 T in the first test and 6.36 K for both legs after 1000 electromagnetic load cycles. In the case of TFCN5, T cs was 6.56 K for the left leg and 6.30 K for the right leg at 68 kA/10.78 T in the first test and 6.33 K for the left leg and 6.02 K for the right leg after 1000 electromagnetic load cycles. According to the SULTAN test result, T cs values of TFCN4 and TFCN5 conductor samples meet the ITER acceptance criteria.

A twisted multifilamentary structure is needed for Nb3Sn and NbTi strands to be used in the International Thermonuclear Experimental Reactor (ITER) magnets. The Cable-In-Conduit Conductor (CICC) for the coils of ITER relies on twisted, multifilament, Nb3Sn and NbTi based composite strands. As important parameters for the superconducting strand design and production, the twist pitch length and the direction of strands must meet the requirements according to ITER Procurement Arrangement (PA) and this must be verified. The technical requirements include 15 mm ± 2 mm for twist pitch length and right hand twist for direction. The twist pitch and the twist direction can be measured on straight sections of strand, which is recognized by the repetition of filament bundles or by the angle of the filaments. Several test methods and results were described and compared in the present paper. The accuracy, the uncertainty and the feasibility of different methods were analyzed and recommended measurement methods were proposed for ITER strands verification.