CICC Joint Development and Test for the Test Facility

The superconducting joint of the NbTi Cable-in -conduit Conductor (CICC) has been developed and tested on the magnet test facility at Institute of Plasma Physics, Chinese Academy of Sciences. The CICC is composed of (2NbTi+lCu)x3x3x(6+ltube) strands each with 0.85 mm in diameter, which has been developed for a central solenoid model coil. The effective length of the joint is about 500 mm. There have been two common fabrication modes, one of them is to integrate the 2 CICC terminals with the copper substrate via lead-soldering, and the other is to mechanically compress the above two parts into an integrated unit. In the current range from 2 kA to 10 kA the joint resistance changes slightly. Up to now, 11 TF magnets, a central solenoid model coil, a central solenoid prototype coil, and a large PF model coil of PF large coil have been completed via the latter joint in the test facility.     

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.     

Manufacture of ITER feeder sample conductors

The ITER feeders are the components that connect the ITER magnet systems located inside the main cryostat to the cryogenics, power-supply and control system interfaces outside the cryostat. The feeder busbars rely on the Cable-In-Conduit Conductor (CICC) design concept as all the conductors for the ITER magnet systems. There are two types of busbars for the feeder systems. One is the Main Busbar (MB) for the TF, CS and PF feeders, and the other is the Corrector Busbar (CB) for the CC feeders. The busbar cable is wound from multiple stage sub-cables made with Cu and superconducting strands. The superconducting material is NbTi for the busbar strands of all feeder systems. All Feeder conductors are provided by China. The R&D programs are needed to acquire knowledge on the behavior of such conductors.Since the conductors are new, some full size copper dummy conductors have been produced for the testing of the cabling parameters, definition of automatic TIG welding of seamless jacket section, elaboration of cable insertion and compaction. Then, two short qualification conductor samples (MB and CB) are prepared in ASIPP, and NbTi advanced strands are produced by Western Superconductor Technology (WST).The details of manufacturing procedures for Feeder conductor samples will be described in this paper.     

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.     

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.     

ITER用极向场线圈导体短样分流温度测试

ITER用极向场（PF）线圈CICC导体短样是用西部超导材料科技有限公司提供的NbTi超导股线绕制完成,该股线在不同温度下的临界电流测试性能稳定,符合绕制导体的要求。对PF导体短样在SULTAN实验室进行了测试,经电磁循环通电前后,分流温度无较大改变,导体性能稳定。在考虑了导体自场作用的情况下,导体在5T、50kA运行环境下的分流温度为6.33K,满足ITER规定的要求。     

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.     

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.     

Characterization of industrial NbTi strands at variable field for JT-60SA toroidal field coils

The mission of the JT-60SA Tokamak, to be built in Japan, is to contribute to the early realization of fusion energy by its exploitation in support of the ITER program. JT-60SA project is an important part of the “broader approach” activity as a satellite program for ITER. The toroidal field (TF) coils are a European “in kind” contribution and they will partly be built by France. JT-60SA TF coil uses the Cable In Conduit Conductor (CICC) with NbTi superconductor strands. TF conductors will have to operate at 5.7 T, 5 K and at current density of 450 A/mm² with sufficient margins. In the framework of JT-60SA TF coil manufacture, the variable temperature characterization is an important step to select NbTi strand. At an early stage of design, we had to choose the strand with acceptable performances. During the design qualification and validation stage, it is important to qualify strands in conditions close to the operation conditions. The industry has proposed various strands manufactured with different processes. This work and publication examines a strand with an internal CuNi barrier, which is expected to lead to better current distribution between strands, by more precise calibration and control of the inter-strand resistance. The strands were tested at the Grenoble High Magnetic Field Laboratory facility. The domain (B, T, J) explored was in the range of 4.5–11 T for the magnetic field intensity, 4.2–6.5 K for the temperature and between 40 A/mm² and 1200 A/mm² for the current density. For each strand, “critical current density” and “current sharing temperature” measurements have been carried out, with a temperature precision of few tens of mK. Once the measurements performed, the fitting parameters (of type J_C = f(B, T)) of each strand have been found, by performing regression analysis. This work will lead to select the strand with the best characteristics. In this paper, we present the results of this measurement task, the data and regression analysis (fits, T_cs, etc.) and the conclusion about the strand choice.     

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...     

The design and the manufacturing process of the superconducting toroidal field magnet system for EAST device

The toroidal field (TF) magnet system of EAST (HT-7U), which consists of 16 superconducting coils enclosed in steel cases, has been manufactured to generate the magnetic field of 3.5 T at the plasma center to maintain plasma in a tokamak configuration with a current up to 1 MA. The TF coils have an approximately D-shape geometry of 2.6 m wide and 4.0 m high, and operate at a maximum field of 5.8 T. The conductor used in the TF coil is NbTi/Cu cable-in conduit (CIC) conductor, and its operating current is 14.3 kA.In March 2006, the first cooling down of the EAST device has been carried out successfully. The total of TF magnet system has been cooled down from room temperature to 4.5 K, and the TF system has been energized up to 8.2 kA with 5 A/s ramp rate. In September 2006, full performances of the TF magnet system have been reached, and the device of EAST has delivered its first plasma. In addition, the TF magnet system has been routinely operated with a current maintained constant on a whole day basis, for a preliminary program of more than 500 shots.In this paper, the main parts of the design, developmental tests, and the fabrication and assembly of TF coils are described in detail.     

Advanced thermal-hydraulic analysis of the ITER TF magnets performed with VENECIA code for 15 MA reference scenario

Normal operation of the ITER TF coils at 15 MA reference scenario is simulated with the use of the VENECIA code. The developed numerical model adopts a full scale quasi 3D approach for thermal hydraulic and thermal diffusion analysis of TF coils at the reference scenario with greatly variable heat loads from nuclear heating and Eddy/AC losses. The model implements latest heat load specifications and corrective changes in design of TFWP, TF case and their cryogenic circuits. For the first time the primary auxiliary cryogenic boxes (ACBs) are included in a common model to provide for the forced-flow cooling of the TF winding, TF case together with CS/OIS structures and PF supports.     

Status of European manufacture of Toroidal Field conductor and strand for JT-60SA project

In the framework of the JT-60SA project, part of the Broader Approach (BA) agreement, EURATOM provides to Japan, the Toroidal Field (TF) magnet system, consisting of 18 superconducting coils. The procurement of the conductor for the TF coils is managed by Fusion for Energy, acting as EU representative in the BA agreement. The TF conductor procurement is split into two contracts, one dedicated to the production of Niobium Titanium (NbTi) and Cu strand and the other to TF conductor production through strand cabling and cable jacketing operations.The TF conductor is a rectangular-shaped cable-in-conduit conductor formed by 486 (0.81 mm diameter) strands (2/3 NbTi–1/3 Cu) wrapped in a stainless steel foil and embedded into a stainless steel jacket.The 18 TF coils require (including spares) 115 ‘Unit Lengths’ (UL) of such conductor, each 240 m long for a total of about 28 km. Correspondingly about 10,000 km for NbTi and 5000 km for Cu strand are produced.The Japanese company Furukawa Electric Co. (FEC) is in charge of TF strand manufacture while the Italian company Italian Consortium for Applied Superconductivity (ICAS) is in charge of cabling and jacketing of TF conductor ULs. In the paper, we provide information on the production stages presently achieved in TF strand and conductor contracts.     

Performance evaluation and analysis of ITER poloidal field conductors

The performance evaluation and analysis of PF5 conductor of the ITER Project in China have been performed using the Gandalf code (Bottura [1]). This study focuses on the T_cs and MQE of PF5 conductor with Cu–non Cu ratio of 2.3 NbTi strands from WST. The PF5 conductor samples have been measured in SULTAN at CRPP for evaluating the performance successfully. The measurement results are also presented with the evaluation results in the paper. The evaluation results related to T_cs and MQE are agreed well with the measurement results. The simulation with Gandalf code can predict the performance of PF5 conductor effectively and provide the helpful method for ITER conductor design and analysis.     

Test and Analysis of China's First Short Conductor Sample for ITER Toroidal Field Coils

In the framework of the ITER Qualification Tests, the first China TF conductor sample (CNTF1) was tested at the SULTAN facility. The sample was made of two TF conductor sections manufactured from identical internal stannum strands provided by the Oxford Superconducting Technology Company (OST). In order to evaluate the conductor performance, the current sharing temperature (Tcs) was measured at specified electromagnetic load cycling steps. Both conductor sections of the CNTF1 sample showed identical performance. Tcs was 7.2 K before cycling loading, and 6.9 K even after 950 cycles, without significant degradation, which substantially exceeds the ITER requirement of 5.7 K. The tests of the CNTF1 conductor sample showed that the electromagnetic cyclic load exhibited a negligible effect on the conductor performance. The coupling time constant for AC loss was 214 ms and 71.52ms before and after the cycling load, respectively. The test results of the sample are compared with the strand performance and parameter model analysis.     

Characterization of superconducting wires and cables by X-ray micro-tomography

Due to their mechanical strength and ability to withstand the large electromagnetic force applied to the superconductors in large magnets during excitation, the Cable-in-Conduit-Conductor (CICC) type superconductors will be employed in the next stage of fusion magnets. Here, we discuss the recent results on the application of a non-invasive method for the characterization of CCIC by X-ray micro-tomography (μXCT). The experiments have been carried out on a high resolution X-ray tomograph in INFLPR (http://tomography.inflpr.ro). An open type nanofocus X-ray source with maximum high voltage of 225 kVp at 15–30 W maximum power and multiple targets of W on different windows materials (Be, Al, Cu or diamond) is the main component. X-rays are detected by means of amorphous silicon flat panel sensor in the cone-beam configuration and high-energy efficient line sensor based on individual scintillators in the fan-beam scanning configuration. The quality of tomographic images (≈40 μm space resolution) allowed the majority of strands of analyzed CICC samples to be fully reconstructed along the investigated segment (up to 300 mm long). Our method provides: (i) local and global void fractions (over a 300 mm length of the sample), (ii) void homogeneity factor as the ratio between void space surface and perimeter and (iii) twist pitch angle of individual strands and its distribution in 3D. It can be used to investigate superconducting CICC during their manufacture, installation or after service inspection, for purposes of QA, characterization or development.     

Stability and design criterion for cable-in-conduit-conductors with a broad transition to normal state

Stability criterion in cable-in-conduit conductors (CICC) is often associated with heat removal rate higher than heat generated in the normal zone, which requires low current density in the strands. We show that this criterion is not a mandatory requirement for serviceability of CICC and that CICC may work reliably at higher current densities. In conditions of limited and well defined perturbations, sufficient stability is provided not by a large amount of copper and high transient heat transfer, but by a smooth transition to the normal state and easy current redistribution. A strand parameter space in terms of I_c and N-value meeting CICC requirements for stability, limited heat generation, and minimum temperature margin is proposed and discussed. The theory predictions are compared with known experimental data on CICC.     

Concept design of CFETR superconducting magnet system based on different maintenance ports

CFETR which stands for “China Fusion Engineering Test Reactor” is a new tokamak device. Its magnet system includes the Toroidal Field (TF) winding, Center solenoid winding (CS) and Poloidal Field (PF) winding. The main goal of the project is to build a fusion engineering Tokamak reactor with its fusion power is 50–200 MW and should be self-sufficiency by blanket.In order to ensure the maintenance ports design and maintenance method, this article discussed the concept design of the magnet system based on different maintenance port cases. The paper detailed studied the magnet system of CFETR including the electromagnetic analysis and parameters for TF (CS)PF. Besides, the volt-seconds of ohmic field are presented as detailed as possible in this paper. In addition, the calculations and optimizations of equilibrium field which should guarantee the plasma discharge of single null shape is carried out. The design work reported here illustrates that the present maintenance ports will not have a great impact on the design of the magnet system. The concept design of the magnet system can meet the requirement of the physical target.     

AC Loss of ITER Feeder Busbar in 15MA Plasma Current Reference Scenario

Superconducting feeder busbar is a crucial component of the International Thermonuclear Experimental Reactor feeder system. AC losses including hysteresis loss and coupling-current losses (10 cycles and 1000 cycles) of the toroidal field feeder main busbar are described and calculated in detail in 15MA plasma current reference scenario in this paper. The critical current density is taken independent of magnetic field, and then the hystersis loss is evaluated; the calculated values of time constant of strands and sub-cables and the test values are approximate, and then the coupling losses are evaluted, too.     

Anomalies on V–I characteristic of NbTi cable-in-conduit conductors

Two kinds of anomaly, observed in the Volt–Ampere Characteristic (VAC) of large NbTi cable-in-conduit conductors (CICC) are discussed. In one case, the wavy behavior of the VAC close to the current sharing range is explained with oscillations of the cooling temperature of the order of 10–30 mK. A simulation of periodic temperature variations is done to reproduce the experimental behavior. In another case, voltage spikes in the VAC are correlated with saw-tooth signals from Hall sensors monitoring the conductor self-field, suggesting the occurrence of local quenches and recovery, with local current re-distribution. The analysis of the two kinds of anomaly, with their own signature on the VAC, provides valuable diagnostic tools for the interpretation of large size CICC test results.