The Test Facility for the EAST Superconducting Magnets

A large facility for testing superconducting magnets has been in operation at the Institute of Plasma Physics of the Chinese Academy of Sciences since the completion of its construction that began in 1999. A helium refrigerator is used to cool the magnets and liquefy helium which can provide 3.8 K-4.5 K, 1.8 bar-5 bar, 20g/s-40g/s supercritical helium for the coils or a 150 L/h liquefying helium capacity. Other major parts include a large vacuum vessel （3.5 m in diameter and 6.1 m in height） with a liquid nitrogen temperature shield, two pairs of current lead, three sets of 14.5 kA-50 kA power supply with a fast dump quench protection circuitry, a data acquisition and control system, a vacuum pumping system, and a gas tightness inspecting devise. The primary goal of the test facility is to test the EAST TF and PF magnets in relation to their electromagnetic, stability, thermal, hydraulic, and mechanical performance. The construction of this facility was completed in 2002, followed by a series of systematic coil testing. By now ten TF magnets, a central solenoid model coil, a central solenoid prototype coil, and a model coil of the PF large coil have been successfully tested in the facility.     

The AC losses measurement and analysis of superconducting NbTi CICC for HT-TU superconducting Tokamak

Superconducting TF and PF coils have been measured in SULTAN test facility. Segregated copper strands are included in four NbTi CICC and this is a technical innovation. Two AC losses measurement methods, calorimetric and electromagnetic methods, have been used in the experiments, and a broad frequency range (from 0.05 Hz to 6 Hz) is covered in sample test. The purpose of this experiment was to investigate AC losses of TF and PF CICC conductor including segregated copper and to check the design of PF and TF CICC coated with different resistive barriers (Pb-30Sn-2Sb and Ni plating on NbTi strands).     

Finite Element Analyses and Instrumentation Layout for Single Coil Testing of TF Coils in HT-7U

The HT-7U tokamak is a magnetically-confined full superconducting fusion device, consisting of superconducting toroidal field (TF) coils and superconducting poloidal field (PF) coils. These coils are wound with cable-in-conductor (CICC) which is based on UNK NbTi wires made in Russian '. A single D-shaped toroidal field magnet coil will be tested for large and expensive magnets systems before assembling them in the toroidal configuration. This paper describes the layout of the instrumentation for a superconducting test facility based on the results of a finite element modeling of the single coil of toroidal magnetic field (TF) coils in HT-7U tokamak device. At the same time, the design of coil support structure in the test facility is particularly discussed in some detail.     

Current Sharing Temperature Test and Simulation with GANDALF Code for ITER PF2 Conductor Sample

Cable-in-conduit conductor （CICC） conductor sample of the PF2 coil for ITER was tested in the SULTAN facility. According to the test results, the CICC conductor...     

Bridge joint fabrication and validation for SST-1 PF coil winding pack

A novel concept of bridge joint for Poloidal field (PF) magnet of SST-1 with damaged winding pack has been realized. This joint has been fabricated on 5th and 6th layers of PF#3T coil winding pack (WP) after validation at 10 kA at liquid helium temperature of 4.2 K in current lead test chamber. The joint resistance of bridge joint was measured ∼1.6 nΩ at flat top DC current of 10 kA. This type of joint could be economically useful for revival of a shorted and damaged WP superconducting PF magnets of Tokamaks. In this paper, details of bridge joint design, fabrication and validations are discussed.     

Present Status of the KSTAR Superconducting Magnet System Development

The mission of Korea Superconducting Tokamak Advanced Research (KSTAR) project is to develop an advanced steady-state superconducting tokamak for establishing a scientific and technological basis for an attractive fusion reactor. Because one of the KSTAR mission is to achieve a steady-state operation, the use of superconducting coils is an obvious choice for the magnet system. The KSTAR superconducting magnet system consists of 16 Toroidal Field (TF) coils and 14 Poloidal Field (PF) coils. Internally-cooled Cable-In-Conduit Conductors (CICC) are put into use in both the TF and PF coil systems. The TF coil system provides a field of 3.5 T at the plasma center and the PF coil system is able to provide a flux swing of 17 V-sec. The major achievement in KSTAR magnet-system development includes the development of CICC,the development of a full-size TF model coil, the development of a coil system for background magnetic-field generation , the construction of a large-scale superconducting magnet and CICC test facility. TF and PF coils are in the stage of fabrication to pave the way for the scheduled completion of KSTAR by the end of 2006.     

绕制预应力对MICE超导耦合磁体冷质量应力状态的影响

超导耦合螺线管磁体为μ介子离子化冷却实验装置(MICE)中的关键设备,其线圈内径1500mm,长度285mm,采用截面1.65mm×1.00mm的NbTi复合超导线,励磁到210A时,峰值磁场可达7.4T。在降温和励磁过程中,为减小导线窜动而导致失超,线圈绕制过程中需对导线和紧固带施加预应力。本文根据组合筒理论,得出了绕制过程中线圈和紧固带的预应力与冷质量内部应力分量的关系。采用有限元方法对线圈绕制、冷却和励磁3个连续过程进行动态仿真,分别分析了导线和紧固带绕制预应力的变化对冷质量内部各主要应力峰值的影响,得出线圈和紧固带绕制时满足磁体稳定性和结构安全的预应力优化结果,为MICE超导耦合磁体的研制及其他类似大直径、多层的超导螺线管磁体绕制提供理论依据。     

The Investigation on Welding Processes for SUS316LN Tubes Used in Superconducting Magnetic System of EAST

The force flow cooled superconducting cable-in-conduit conductor （CICC） is used in both of EAST toroidal field （TF） and poloidal field （PF） coils. The conductor consists of multi-stage NbTi superconducting cable and 1.5 mm thick square jacket. The cable is pulled through in a thin wall circular jacket and then compacted to square cross-section conductor. The jacket material is SUS316LN austenitic stainless steel seamless tubes （about 10 m each）, which is assembled by butt-welding up to 600 m. The results of the welding procedure investigation and quality assurance procedures carrying out are described in this paper.     

Simulation and test of quench in CICC: The SMES experience

This paper summarizes the work done as part of the U.S. SMES program to simulate quench evolution on the 200 kA CICC developed by the Bechtel Team. As a large-scale CICC with a central tube, this work has led to a number of results applicable to other conductors sharing similarities with the SMES-CICC. The paper presents the evolution in computational models, since 1987 to date, and describes QUIPS, a test intended to validate these computer models. The paper concludes with observations on the directions in the field as perceived by the author.     

HT-7U NbTi 管内电缆导体瞬态稳定性的实验研究与分析

在SULTAN 测试设备上进行了含分离铜股线CICC瞬态稳定性的实验研究,应用脉冲场对样品的脉冲场区域(390 mm)进行感应加热,发现设计的含分离铜股线CICC能够经受住很大的瞬态磁扰动,分析了这个现象的原因,并就股线上的电阻层对稳定性的影响进行了分析,对4个CICC导体的稳定性差异进行分析和稳定性裕度的理论计算,由理论计算值和实验测量值进行比较分析,为HT-7U纵场和极向场NbTi CICC的选择提供实验依据.     

Construction of a 2kW／4K Helium Refrigerator for HT—7U

Superconducting magnets of toroidal field (TF) and poloidal field(PF) of HT-7U tokamak are all made of NbTi/Cu Cable-in-Conduit Conductor (ClCC),and cooled with a forced flow supercritical helium of 3.8K.A helium refrigerator with an equivalent capacity of 2kW/r K will be constructed.This paper presents the design of the helium refrigerator process.The thermodynamics of the refrigeration cycle and the refrigerator equipments.     

Power dissipation and energy transfer during testing of the ITER toroidal field model coil

The testing of the ITER toroidal field model coil (TFMC) in the background field of the EURATOM-LCT coil took place in autumn 2002 at the TOSKA facility of the Forschungszentrum Karlsruhe in the framework of the ITER R&D programme. The maximum currents in the two coils, in combined operation, were 16 kA in the LCT coil and 80 kA in the TFMC, respectively. The heat load of both coils, including the eddy current losses in the passive structures and the joule losses due to the joint resistances, was removed by a secondary loop of forced flow supercritical He. About 2% of the stored energy was transferred to the cryogenic system after all the safety discharges of both coils together. Most of the energy (about 98%) was extracted and transferred to the dump resistors of both coils, located outside the vacuum vessel. A computer code, based on the full inductance and resistance matrices, has been developed with SIMULINK™. After validation with experimental data the code has been used to perform circuit analysis and to evaluate the power dissipation and energy transferred to the cryogenic plant and to the external power circuits.     

Fabrication of new joints for SST-1 TF coil winding packs

The Toroidal Field (TF) magnet system of SST-1 has sixteen NbTi/Cu based coils with about one hundred Inter-Pancake (IP) and Inter-Coil (IC) joints. New box type helium leak tight, low DC resistance joints have been designed, fabricated and tested at 5 K temperature and 10 kA DC transport current. The prototype of this joint has been validated in laboratory as well as on spare TF coil winding pack. Moreover, the performance of these joints has been realised and validated on actual sixteen TF winding packs, the joint resistance of ～0.5 nΩ repeatedly measured on hundreds of IP joints. The quality of terminations and joints was ensured at various stages of fabrication. The quality of joint box material was ensured by visual inspection, chemical analysis, radiography test, ultrasonic test, eddy current test, etc. This paper describes joint design drivers, joint design detail, prototype joint fabrication processes, quality assurance (QA)/quality control (QC) adopted during prototype and actual joint fabrication process, joint resistance measurement on actual TF coils and analysis of measured joint resistance in detail.     

1-D thermal-hydraulic analysis of the high temperature superconducting current leads for the ITER magnet system from 5 K to 300 K

The magnet system of ITER includes high temperature superconducting (HTS) current leads with a maximum current of 68 kA for the toroidal field (TF) coils, 55 kA for the poloidal field (PF)/central solenoid (CS) coils and 10 kA for the control coils (CC), respectively. Although different in terms of size and operative conditions, the ITER HTS current leads have been all designed on the basis of an established concept, which was successfully developed for the LHC at CERN and proven by the so-called 70 kA “demonstrator” lead made by KIT and by the ITER pre-prototypes made by ASIPP in China. A broad R&D campaign has been undertaken by ASIPP and CERN in order to find optimized designs for each component of the leads. Nevertheless, a comprehensive picture of the performance of the entire HTS current leads is not yet available. In this paper, a steady state, full length, thermal-hydraulic 1-D modeling is applied to the study of the three types (TF, PF/CS, CC) of ITER HTS current leads. The results of this predictive analysis are then compared with relevant ITER requirements. It was found that the present design of the HTS current leads will fulfill these specifications.     

How should we test the ITER TF coils?

The international thermonuclear experimental reactor (ITER) toroidal field (TF) magnet system consists of 18 superconducting coils using a 68 kA Nb₃Sn conductor. In order to guarantee the performances of these coils prior to their installation, the test of at least one prototype coil at liquid helium temperature and full current is required. The test of all coils in the two-coil test configuration, with successive charging of each coil to nominal current is recommended. This requires a large test facility.     

Numerical and experimental analysis of AC loss for CFETR CS model coil

The central solenoid (CS) is an important com ponent of China Fusion Engineering Test Reactor,for producing,forming and stabilizing plasma in the superconducting tokamak.It is a complicated work to design and manufacture the large superconducting CS magnet,so it is meaningful to design a central solenoid model coil (CSMC) and analyze its electromagnetic properties in advance.In this paper,the structure,design parameters and magnetic field distribution of the CS model coil are dis cussed.The peak power of radial and axial turn conductors and time bucket loss are analyzed by using piecewise-linear method.The CSMC AC loss with different Nb3Sn CICCs and AC loss of ITER CS coil are compared.The special electrometric method to measure AC loss of the CS model coil for fu ture reference is presented.     

Modeling AC losses in the ITER NbTi poloidal field full size joint sample (PF-FSJS) using the THELMA code

A first validation of the full version of the thermal–hydraulic electromagnetic (THELMA) code, developed for the analysis of transients in the cable-in-conduit conductors and coils relevant for the International Thermonuclear Experimental Reactor (ITER) is presented here. THELMA includes electromagnetic models of the cable joints and terminations (lumped parameter) and of the conductor (distributed parameter), while for the thermal–hydraulics of the helium coolant it includes a compressible 1D flow model. The AC losses induced by a pulsed coil in the NbTi poloidal field full size joint sample (PF-FSJS) right leg conductor, tested in 2002 at the Sultan facility in Villigen (CH), are considered as test bed for this exercise. The computed energy deposition and evolution of the temperature downstream of the heated zone are in good agreement with the measured values. However, the inter-bundle electrical conductances needed in input by the code are compatible with measured values only when a sufficiently refined model is used in the cable cross-section.     

The first performance of the facility for testing ITER CC conductor short sample

A new facility had been set up to test the low temperature properties of the short sample of the small-size cable-in-conduit conductor (CICC). The facility consisted of the background magnet which could provide 7 T centric magnetic field, a 50 kA superconducting transformer which provided sample current, a 500 W/4.5 K helium refrigerator which produced both the liquid helium (LHe) and supercritical helium (SHe). An ITER CC conductor short sample was prepared and measured in this testing system. T_cs of 7.02 K (@4.1 T, 10 kA) and I_c of 8.9 kA (@4.1 T, 7.06 K) were measured.     

BRISOL谱仪关键技术研究

HI-13串列加速器升级工程在线同位素分离器（BRISOL）需对同质异位素进行分辨,谱仪设计质量分辨率为20 000,是很高的技术指标,对离子源、高压、分析磁铁、四极透镜等设备均有很大的挑战。本文详细介绍了BRISOL谱仪关键技术及其测试结果。能散对谱仪的质量分辨率影响较大,BRISOL谱仪设计采用异能大小铁结构消除能量色散。离子源采用表面离子源,并采用三电极引出系统,中间电极电压可调用以优化束流品质,优化后离子束RMS发射度好于3.8 πmm•mrad。分析磁铁采用表面线圈进行磁场垫补,垫补后积分场均匀性好于3.5×10^-5。为修正像差,大分析磁铁安装了β线圈和γ线圈,同时,在分析磁铁前后共设置了4台电六极透镜。     

Design of the dummy coil for magnet power supply

It is necessary to test it on a dummy coil, before using a magnet power supply (MPS) to energize a Poloidal Field (PF) coil in the Korea Superconducting Tokamak Advanced Research (KSTAR) device. The dummy coil should accept the same large current from the MPS as the PF coil and be within the capability of the utilities located at the KSTAR site. Therefore a coil design based on the characteristics of the MPS and other restrictive conditions needed to be made. There are three requirements to be met in the design: an electrical requirement, a structural requirement, and a water cooling requirement. The electrical requirement was that the coil should have an inductance of 40 mH. For the structural requirement, the material should be non magnetic. The coil support structure and water cooling manifold were made of SUS 304. The water cooling requirement was that there should be sufficient flow rate so that the temperature rise ΔT should not exceed 12 °C for operation at 12.5 kA for 5 min. Square cross-section hollow conductor with dimensions of 38.1 mm × 38.1 mm was used with a 25.4 mm center hole for cooling water. However, as a result of tests, it was found that the electrical and structural requirements were satisfied but that the water cooling was over designed. It is imperative that the verification will be redone for a test with 12.5 kA for 5 min.