Plasma Physics Chinese Academy Research Articles

Comparison of Critical Current Versus Axial Strain Measurements on Internal Tin <inline-formula> <tex-math notation="TeX">$\hbox{Nb}_{3}\hbox{Sn}$</tex-math></inline-formula> Strand at ASIPP and University of Twente

In the last decades, various methods have been used to characterize superconducting behavior versus axial strain. In cooperation with the University of Twente (UT), the Institute of Plasma Physics Chinese Academy of Sciences (ASIPP) built a facility practically similar to the so-called “Pacman” spring device. The facility allows critical current measurements of an Nb3Sn strand at different temperature, magnetic field and applied axial strain. In the process of validating the new setup, an ITER Internal-Tin (IT) strand, which has also been measured at the UT, is tested with the ASIPP Pacman. The ITER IT wire has an irreversible strain limit around +0.3%. At ASIPP, the strand was characterized in a magnetic field from 10 to 13 T and applied axial strain from -0.7% to +0.5%. Similarities of the data obtained (IC(e), IC(B) and irreversible strain limit) from the two institutes were remarkable.

Thermal hydraulic analysis of the HECLIC blanket breeder unit for CFETR

Summary China Fusion Engineering Test Reactor (CFETR) is proposed by the China National Integration Design Group. A helium-cooled lithium ceramic (HECLIC) blanket for CFETR has been designed by Institute of Plasma Physics Chinese Academy of Sciences (ASIPP) and University of Science and Technology of China (USTC). The HECLIC blanket includes the tungsten armor, first wall (FW), breeder units (BUs), caps, stiffening grids and back plates. The BU consists of the beryllium pebble beds, lithium ceramic pebble beds and cooling plates. The stiffening grid reinforces the blanket structure and also cools the BU together with cooling plates. Thermal hydraulic analysis has been performed to assess the heat transfer coefficient (HTC) and temperature distribution. The results indicate that the velocity is fairly uniform in the cooling loops, and the HTC is high enough to remove the heat timely. The maximum temperatures are within the design temperature limits. Besides, optimization has been done to improve the layout of the cooling channels. In addition, thermal mechanical analysis has been carried out. The maximum stress can satisfy design limits, and it proves the feasibility of the design. Copyright © 2014 John Wiley & Sons, Ltd.

Nb3Sn Strand Verification for ITER TF Conductors of CNDA

China is charged with around 7.5% of the ITER toroidal field (TF) conductor fabrication. In total about 6000 km of Nb3Sn strands need to be supplied for Chinese Domestic Agency (CNDA), and now the TF procurement arrangement (PA) for CNDA is in the last phase, i.e., production. According to the PA, the Domestic Agency (DA) has to verify on a regular basis the room and low temperature measurements carried out by the strand suppliers. The verifications shall be performed by the reference laboratories on the strand verification samples adjacent to the samples used by the strand suppliers for critical current measurements. As the reference laboratory of CNDA, the superconducting strand test laboratory of Institute of Plasma Physics Chinese Academy of Sciences (ASIPP) is undertaking the task of Nb3Sn strand verification. The verifications include diameter, chromium plating thickness, twist pitch direction and length, copper-to-non-copper volume ratio, residual resistance ratio (RRR), critical current (IC) and resistive transition index (n), and hysteresis loss. The verification level was based on the requirements of the ITER PA. From the results, the supplied strands for the TF conductors fulfill well the requirements of the ITER PA.

Investigation and analysis on ITER in-vessel coils’ raw-materials

The ITER in-vessel coils (IVCs) consist of 27 coils edge localized modes (ELM) and 2 coils vertical stabilization (VS) which are all mounted on the vacuum vessel wall behind the shield modules. The IVCs design and manufacturing work is being conducted in between Institute of Plasma Physics Chinese Academy of Sciences (ASIPP) and Princeton Plasma Physics Laboratory (PPPL). Because the position of ELM and VS coils is close and face to the plasma, the IVCs must undergo a severe environment, such as the high dose of radiation and high operation temperature, thus the conventional electrical insulation materials cannot be used. And the technology of “Stainless Steel Jacketed Mineral Insulated Conductor” (SSMIC) is deemed as the best choice to provide the necessary radiation resistance and compatibility strength in ITER's vacuum vessel. While mineral insulated conductor technology is not new, and is similar to the mineral insulated cable used in industrial. Some difficulties still need to be solved, such as searching for the proper raw-materials to make sure that the conductor have the properties of high current carrying capability, the necessary radiation resistance, the proper strength, at the same time, it must be come true in manufacture technology. This paper described the analysis of the materials for VS and ELM coil conductor.

Manufacturing of the ITER TF Conductors in China

The toroidal field (TF) coil conductor is a cable-in-conduit conductor made up of superconducting, Nb3Sn-based strands mixed with pure copper strands. The strands are assembled in a multistage cable around an open central spiral. The cable and its spiral are inserted into a stainless steel jacket. The jacket provides the helium confinement and must conform to the leak tightness standard defined herein. A technological complex has been developed at the Institute of Plasma Physics Chinese Academy of Sciences for serial production of ITER TF conductors is described. This technology was developed as a result of performing some R&D programs on selecting cabling parameters, welding technology, elaboration of cable insertion, compaction, and winding process. One qualification dummy TF conductor (760 m) and superconducting conductor (110 m) were successfully made. A new manufacturing complex was developed within the Institute of Plasma Physics Chinese Academy of Sciences for realization of this technology. The details of manufacturing procedures are described in this paper. It consists of a cabling workshop, and jacketing line equipped with full production machinery and test devices.

Qualification Phase of Key Technologies for ITER Correction Coils

The ITER Magnet system [1] consists of four main coils sub-systems: 18 Toroidal Field Coils (TF-coil), a Central Solenoid (CS), 6 Poloidal Field Coils (PF-coil) and 18 Correction Coils (EFCC). The main contract of the EFCCs supply is awarded to the Institute of Plasma Physics Chinese Academy of Sciences (ASIPP) by the Chinese Domestic Agency (CNDA). According to the pre-qualification program, ASIPP is implementing the procurement phase to qualify and validate key technologies and manufacturing methods. The Correction coils qualification activities are conducted within the framework of the procurement arrangement set up between the ITER Organization and CNDA. The paper describes the CC development including first results of the coils winding qualification trials and, qualification of a S-Glass fiber-polyimide based insulation system The CC casing assembly process and the first results of the welding trials are reported. The weld qualification, according to ASTM for 316LN austenitic steel is reported in terms of fracture toughness, fatigue crack growth, and tensile property at 4 K. The Vacuum Pressure Impregnation of CC short mock-up, with low viscosity bisphenol-F (DGEBF) epoxy resin, aims to optimization of the curing and insulation mechanical properties.

Some technology issues for the general assembly of EAST superconducting Tokamak

The construction of experimental advanced superconducting Tokamak (EAST) was completed at the Institute of Plasma Physics of Chinese Academy of Sciences (CASIPP) in January of 2006 after challenging techniques over the period of 8 years. The followed cooling-down test was also successfully performed and all superconducting magnets including two current leads and seven bus lines entered the cryogenic superconducting state smoothly. Total 260 shots have been fired. The longest TF current duration was 5000 s and the highest TF current was 8200 A (2T). All the testing results showed that the EAST machine and its sub-systems have been successfully built. EAST machine is a fully superconducting Tokamak. It consists of more 20,000 parts and possesses the configuration of 7600 mm in diameter, 6600 mm in height and the weight of 400 tonnes. The complicated general assembly of EAST began in July of 2003 and nearly took two and half years. During the process, not only a series of common difficulties for a large Tokamak but also many technique problems for a superconducting Tokamak itself needed to be solved properly. This paper emphasizes on some key technology issues that were faced and solved in EAST assembly process including optimization of assembly procedure, survey and alignment, ensuring of thermal insulation gap, installation of unique parts, treatment of electric insulation, improvement on heat conduction, control on deformation of vacuum vessel, etc. Besides, the quality control (QC), which made important contributions to the general assembly is also described.

The batch production for superconducting magnet coils of EAST (HT-7U)

The experimental advanced superconducting tokamak (EAST) is being constructed at the Institute of Plasma Physics of Chinese Academy of Sciences (CASIPP). The EAST project, approved by the Chinese government as a national mega project of science research is a fully superconducting tokamak. The most key component for EAST is the superconducting magnet coils (SMCs), which consists of 16 toroidal field coils (TFCs) and 14 poloidal field coils (PFCs). In 2003, three prototypes, one TFC and two PFCs, were successfully completed and passed a series of cryogenic tests. Batch production, needed for the SMCs has begun at CASIPP since 2002. Up to now, all 58 CIC conductors with a total length of 32 km, 12 TFCs out of 16 and 10 PFCs out of 14 have been fabricated. This paper emphasizes on the various technology issues that must be faced and solved for four R&D lines of SMCs after transforming to batch production. Quality control methods in process are also described.

A 600-m Jacketing Line of Cable-in-conduit Conductor (CICC) and the First 600-m Dummy Cable

A cable-in-conduit conductor (CICC) production line was designed and constructed in Institute of Plasma Physics of Chinese Academy of Sciences (IPPCAS) by the end of 2000. It can produce a length of 600 meters and three kinds of sections of 20.8±0.1 × 20.8±0.1, 20.4±0.1 × 20.4±0.1 and 18.6 ± 0.1 × 18.6 ± 0.1 mm2. If the rollers of the shaping machine are changed, it can also produce other sizes of CICCs. So-called inserting-cable technology is adopted in this production line, where the procedures consist of tube pre-treatment (cleaning, pressure and leakage testing, end cutting), conduits butt-welding, six kinds of quality checking (endoscopy, dye penetration, pressure control, leakage testing, ultrasonic inspection and x-ray testing), cable inserting, shaping (compacting & squaring), pre-bending & winding and final checking. Now all the instruments and facilities required for these technologies have been installed and got ready. Some key technologies have been explored and good results obtained. Some short samples were produced and a 600 meters long sample was made out in August, 2001.

Preliminary Engineering Design of Toroidal Field Magnet System for Superconducting Tokamak HT-7U

HT-7U is a large fusion experimental device. It will be built in the Institute of Plasma Physics of Chinese Academy of Sciences. The mission of HT-7U is to develop the scientific basis for a continuously operating tokamak fusion reactor. This paper describes only a toroidal field (TF) superconducting magnet system of HT-7U. In this paper, design criteria of conductor and stability analysis, coil winding and support structure design of magnet system, mechanical calculation and stress analysis, heat load evaluation are given.

Preliminary engineering design of toroidal field magnet system for superconducting tokamak HT-7U

HT-7U is a large fusion experimental device. It will be built at the Institute of Plasma Physics of Chinese Academy of Sciences. The mission of HT-7U is to develop the scientific basis for a continuously operating tokamak fusion reactor. This paper describes only the toroidal field (TF) superconducting magnet system of HT-7U. In this paper, design criteria of conductor and stability analysis, coil winding and support structure design of magnet system, mechanical calculation, stress analysis and heat load evaluation are given.