北京正负电子对撞机重大改造工程中超导磁体电流引线设计

北京正负电子对撞机重大改造工程（BEPCⅡ）中超导聚焦四极磁体(SCQ)共有6对电流引线，输送4种不同大小的电流。超导探测器磁体(SSM)由1对4000A的电流引线输送电流。本文为SCQ和SSM两个超导磁体设计多层套管结构的电流引线。引线通过在低温端增加大质量铜座的方法来延长当冷却氦气消失时低温端温度上升到超导导线失超温度的时间。给出了多层套管结构电流引线稳态与非稳态大型CFD软件Fluent6.0数值模拟结果。     

变负荷电流引线的设计

超导核聚变实验装置(EAST)的极向场超导磁体常常运行在空载、变电流条件下,将这种电流引线设计成过载电流引线可以进一步降低低温系统的热负荷。本文计算了不同材料制作的电流引线在额定电流和过载情况下的漏热、温度分布等参数,在分析计算的基础上给出了制造电流引线时的选材原则以及过载运行的条件。     

ITER10kA电流引线高温超导段研究

国际热核实验反应堆(ITER)高温超导电流引线（HTSCL）的特点是不仅电流容量大,且安全性要求非常高,高温超导段是HTSCL的关键部件。本文论述了ITER10kA电流引线高温超导叠和超导组件的真空钎焊工艺,分析了高温超导段漏热,并对高温超导段漏热和电流引线在10kA下的安全性参数进行了测试。结果表明,电流引线不仅漏热小,且安全裕度大,满足ITER设计要求。     

ITER 10 kA高温超导电流引线试验件设计与测试

本文介绍了10 kA电流引线试验件的设计优化,主要描述了针对ITER所用与要求而设计的10 kA电流引线的设计结构与加工,电子束焊接首次应用于电流引线的焊接工艺。说明了此电流引线试验件的测试性能;对目前的工艺与测试结果提出并讨论了可能的改进方案。     

用于4mm回旋管的超导磁体

描述了一个4mm回旋管用的超导磁体,它可在谐振腔区提供3T的均匀磁场,均匀区长6cm,均匀度优于±0.16%,均匀磁场又可调节成梯度场,梯度△B/B_o≥8.5%。阴极区磁场均匀区长4cm,均匀度±(1.0～1.8)%,压缩比B_k/B_o可在1/6～1/15范围内调节。采用四台电源串联供电。磁场位形调节好后,磁体可实现闭环运行。磁体将与回旋管配合在HL-1装置上进行ECRH物理实验。     

用于超导磁体的P1WM变换器的无振荡控制法

电流型脉冲宽度调制(PWM)变换器被认为是下一代托卡马克装置超导磁体供电的选择之一。但是，这类PWM变换路有一个问题，即在AC侧波滤波器中电感电温和电容电压的瞬时振荡，它的出现是因为当DC输出电流参考值迅速变化时滤波器的LC共振。本文提出了基于无振荡控制原理的PWM控制法，用以抑制瞬时振荡。使用这一方法的变换器，在理想情况一个采样周期就能够控制电感电温和电容电压。另外，还讨论了控制中的空挡时间的作用。这一新提出的方法的性能通过数值电路模拟予以说明。     

8mm回旋管用超导磁体系统的研制

本文描述了一个8mm回旋管用的超导磁体系统的研制。它的主磁场线圈中心轴磁场达到2.14T,均匀区长度为14.5cm,均匀度±3.84‰,利用超导开关可以闭环运行。副磁场均匀区长度为4.5cm,均匀度±1.43‰。杜瓦具有台阶型室温通道,上端直径74mm,下端直径140mm。杜瓦是直筒型的,装有超导磁体时液氦蒸发量为4L/h。     

8mm回旋管用超导磁体系统的研制

本文描述了一个8mm回旋管用的超导磁体系统的研制。它的主磁场线圈中心轴磁场达到2.14T,均匀区长度为14.5cm,均匀度±3.84‰,利用超导开关可以闭环运行。副磁场均匀区长度为4.5cm,均匀度±1.43％。杜瓦具有台阶型室温通道,上端直径74mm,下端直径140mm。杜瓦是直筒型的,装有超导磁体时液氦蒸发量为4L/h。     

脉冲强磁体的概念性设计和行为仿真

本文对一个内径为10 mm、能产生60T场强、持续脉冲时间为10 ms的强磁体的相关参数进行了设计计算,并对该强磁体在不同脉冲电源条件下的电特性行为进行了仿真.     

Computational thermal-fluid dynamics analysis of the laminar flow regime in the meander flow geometry characterizing the heat exchanger used in high temperature superconducting current leads

The Karlsruhe Institute of Technology and the Politecnico di Torino have developed and validated a computational thermal-fluid dynamics (CtFD) strategy for the systematic analysis of the thermal-hydraulics inside the meander flow heat exchanger used in high-temperature superconducting current leads for fusion applications. In the recent past, the application of this CtFD technique has shown that some operating conditions occurring in these devices may not reach the turbulent regime region. With that motivation, the CtFD analysis of the helium thermal-fluid dynamics inside different meander flow geometries is extended here to the laminar flow regime. Our first aim is to clarify under which operative conditions the flow regime can be considered laminar and how the pressure drop as well as the heat transfer are related to the geometrical parameters and to the flow conditions. From the results of this analysis, correlations for the pressure drop and for the heat transfer coefficient in the meander flow geometry have been derived, which are applicable with good accuracy to the design of meander flow heat exchangers over a broad range of geometrical parameters.     

A two-fluid code for the thermohydraulic transient analysis of CICC superconducting magnets

We present here a finite element computer model (Mithrandir) for the transient thermohydraulics of compressible helium in a Cable-In-Conduit Conductor (CICC) with central cooling hole, as presently envisaged for superconducting magnets of the International Thermonuclear Experimental Reactor (ITER). In the model the He in the hole and that in the cable bundle are treated as separate fluids, each characterized by its own flow and thermodynamic properties, coupled by exchanges of mass, momentum and energy. Results for the simulation of a quench both with and without a wall delimiting the central cooling hole are discussed. Time and space convergence of the code are demonstrated numerically.     

The structural design of superconducting magnets for the large coil program

Fusion reactor designs based on magnetic confinement will require the use of superconducting magnets to make them economically viable. For a tokamak fusion reactor; large magnetic field coils are required to produce a toroidal magnetic confinement volume. Although superconductors have been used for approximately 20 years, several requirements for their application in fusion reactors are beyond demonstrated technology in existing magnets. The Large Coil Program (LCP) is a research, development, and demonstration effort specifically for the advancement of the technologies involved in the production of large superconducting magnets. This paper presents a review of the status of the structural designs, analysis methods, and verification tests being performed by the participating LCP design teams in the US, Switzerland, Japan, and the Federal Republic of Germany. The significant structural mechanics concerns that are being investigated with the LCP are presented.     

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.     

高性能微电流集成放大器的设计

介绍了用ICL7650运算放大器设计高性能微电流集成放大器的方法,阐述了电路的工作原理和提高放大器性能的措施,给出了实际应用的实验结果。该放大器测量范围10^-9～10^-12A,自换量程,自动校零,测量准确,工作稳定。     

Bootstrap current fraction scaling for a tokamak reactor design study

We have derived new bootstrap current fraction scalings for systems codes by solving the Hirshman–Sigmar model, which is valid for arbitrary aspect ratios and collision conditions. The bootstrap current density calculation module in the ACCOME code was used with the matrix inversion method without the large aspect ratio assumption. Nine self-consistent MHD equilibria, which cover conventional, advanced and spherical tokamaks with normal or reversed shear, were constructed using numerical calculations in order to compare the bootstrap current fraction values with those of the new model and all six existing models. The Wilson formula successfully predicted the bootstrap current fraction, but it requires current density profile index for the calculation. The new scaling formulas and IPDG accurately estimated the bootstrap current fraction for the normal and weakly reversed shear tokamaks, regardless of the aspect ratio. However, none of the existing models except the Wilson formula can accurately estimate the bootstrap current fraction for the reversed shear tokamaks, which is promising for the advanced tokamak operation mode.     

An initial study of demountable high-temperature superconducting toroidal field magnets for the Vulcan tokamak conceptual design

Recent developments have made it possible to consider high-temperature superconductor (HTS) for the design of tokamak toroidal field (TF) magnet systems, potentially influencing the overall design and maintenance scheme of magnetic fusion energy devices. Initial assessments of the engineering challenges and cryogenic-dependent cost and parameters of a demountable, HTS TF magnet system have been carried out using the Vulcan tokamak conceptual design (R = 1.2 m, a = 0.3 m, B₀ = 7 T) as a baseline. Jointed at the midplane to allow vertical removal of the primary vacuum vessel and routine maintenance of core components, structural D-shaped steel support cases provide cryogenic cooling for internally routed YBCO superconducting cables. The cables are constructed by layering ～50 μm thick commercially available YBCO tape, and the interlocking steel support cases self align during assembly to form internal resistive joints between YBCO cables. It is found that designing the TF magnet system for operation between 10 K and 20 K minimizes the total capital and operating cost. Since YBCO is radiation-sensitive, Monte Carlo simulation is used to study advanced shielding materials compatible with the small size of Vulcan. An adequate shield is determined to be 10 cm of zirconium borohydride, which reduces the nuclear heating of the TF coils by a factor of 11.5 and increases the YBCO tape lifetime from two calendar years in the unshielded case to 42 calendar years in the shielded case. Although this initial study presents a plausible conceptual design, future engineering work will be required to develop realistic design solutions for the TF joints, support structure, and cryogenic system.