230 MeV质子回旋加速器超导线圈力学分析及固化工艺研究

本文对230 MeV质子回旋加速器超导线圈进行了力学分析,对其在绕制、降温、励磁等不同工艺条件下的强度进行了校核,并对其固化工艺进行了研究。通过7次绕制与浸渍实验,得到满足超导线圈固化要求的主要工艺参数。根据以上研究结果制造的超导线圈在励磁过程中没有失超,验证了其具有良好的结构稳定性。     

用于钠泵的螺管型超导磁体设计与实验

本文对钠金属电磁泵的螺管型超导磁体部分进行了详细设计和实验研究.磁体设计指标为水平室温孔长度600mm,直径160mm,中心场磁感应强度5T.对超导磁体所用的超导线进行了短样测试,当温度为4.2K、磁感应强度为5T时,临界电流为464A,大于工作电流.磁体经过绕制、真空压力浸渍工艺后进行10次失超锻炼,在失超电流97 A的电流下测试,磁体中心磁感应强度达到4.66T.若进一步降低励磁速率,磁体中心磁感应强度预计可以达到5T左右,可以满足钠金属电磁泵的要求.     

BEPCⅡ超导插入四极磁体冷却流程的数值分析

按北京正负电子对撞机重大改造工程（BEPCⅡ）计划，为压缩束团尺寸、提高探测器的分辨率以及粒子识别能力，在南对撞区分别安装1对强聚焦超导插入四极磁体（SCQ）和1台超导螺线管探测器磁体（SSM）。本文针对1对超导插入四极磁体的冷却，采用数值模拟的方法给出了SCQ磁体分别采用超临界氦流和过冷氦流两种冷却方式下冷却流程的热力参数，通过对模拟结果的分析，给出了适合该超导插入四极磁体的冷却方式和正常运行的热力参数。还给出了该低温系统关键设备之一的过冷器的设计方法以及设计参数。      

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。     

ITER内部ELM线圈导体残余应力的分析与改善

本文主要介绍使用盲孔法和热时效工艺对辊压缩径成型的ITER内部边界局域模(ELM)线圈导体在弯曲前后其外部铠甲(Inconel 625)及内部铜导体(CuCrZr)的残余应力分布和应力改善情况进行的对比分析和研究。研究表明,导体成型的冷变形工艺会使得导体铠甲及内部铜导体的残余应力增加,相比于成型前残余主应力会增加1倍以上。线圈绕制工艺能够降低线圈外部铠甲、内部铜导体约20%残余应力。热时效处理工艺能够显著缓解绕制后的ELM线圈残余应力,线圈外部铠甲、内部铜导体的残余应力能降低约40%。此外,550℃恒温2h并快冷的热时效工艺不仅降低弯曲ELM导体样品外部铠甲及内部铜导体的残余应力,同时热时效处理能够使得外部铠甲及铜导体上的残余应力均匀化。在均匀化之后,外部铠甲与内部铜导体的平均残余主应力差(σ1,σ2)分别为27.07MPa和15.97 MPa,较导体成型工艺后的平均应力差分别降低了38.98MPa和31.67MPa。     

230 MeV超导回旋加速器超导线圈系统运抵原子能院

正230 MeV超导质子回旋加速器的超导线圈系统采用液氦零挥发的冷却方式,由线圈绕组、低温恒温器、低温制冷机、液氦阀箱,以及真空系统、电源系统及失超保护系统组成。经过近一年半的艰苦攻关,超导线圈系统于2016年12月初在加工厂家完成了装配及过程中的一系列测试,超导磁体经过液氮及液氦淋浴降温,经过不到1周时间线圈降温到4.2K,如图1所示。超导线圈系统一次性成功励磁到场,并完成了不带铁芯的磁场测量与理论设计的磁场吻合。并于2017年的1月完成了源地验收。之后又进     

CFETR纵场超导磁体CICC导体低温冷却设计与分析

纵场磁体中导体低温超导性对整个托卡马克装置磁体系统的稳定具有重要作用。对纵场磁体的低温冷却系统进行热分析,确定线圈排布匝数、纵场磁体中心冷却孔以及盒壁内侧冷却孔的质量流速,并计算液氦在导体中心冷却孔流动时的沿程压降。分析结果表明:同种工况不同线圈排布磁体组件温差值各不相同,经过对比分析,纵场磁体选用154匝线圈排布方案。为确保纵场线圈盒体温度与导体铠甲的温度分别在20 K与5 K范围之内,线圈盒内侧壁面冷却孔质量流速为3.6 g?s~(-1),纵场磁体线圈导体内中心冷却孔质量流速为9.5 g?s~(-1),液氦流动沿程压降为1.72×10~5Pa。     

EAST-NBI偏转磁体线圈水冷能力分析与测试研究

偏转磁体是中性束注入器的关键部件之一,它安装在中性束注入器真空室内部。为适应核聚变研究装置对中性束注入器高能量、长脉冲、稳态运行的要求,对其偏转磁体原先励磁线圈的水冷系统进行了分析,提出增加水冷抽头的方法完成了改进设计,并对改进前后线圈的冷却能力进行了实验测试。测试结果显示,当中性束注入器长脉冲稳态运行时,改进后的水冷结构能及时带走偏转磁体励磁线圈产生的热量;冷却水的进出口水温差约21 oC;偏转磁体线圈导体表面的温度约45 oC;改进设计水冷系统性能得到优化,满足了EAST-NBI高参数、稳态运行的要求。     

脉冲强磁体中应力的有限元分析

提高脉冲磁场强度的主要障碍是磁体中巨大的应力,而脉冲磁体中应力的产生和作用过程比较复杂,要准确地计算出来很困难。采用有限元法对脉冲磁体中的应力进行分析,计算中全面考虑了预应力、电磁力、热应力等情况。通过计算,得出了,一些设计脉冲磁体的基本原则。     

一个适用于穆斯堡尔谱学的超导磁体装置

本文叙述国内研制的第一台穆斯堡尔谱学研究用的超导磁体装置,它包括超导磁体系统、低温系统、电原和保护系统以及样品室变温和控温系统。超导磁体由主线圈和抵消线圈反接串联,最大中心场强7.5T。放射源和吸收体(样品)分别置于抵消线圈和主线圈的中心,带有超导开关的超导磁体系统工作于永久电流状态,具有良好的稳定性。磁体杜瓦用液氮屏和由蒸发的冷氦气冷却的气冷屏结构,蒸发率小于0.2 1/h,杜瓦中心具有可     

Static stress analysis of coupling superconducting solenoid coil assembly for muon ionization cooling experiment

The stresses in the coupling superconducting solenoid coil assembly, which is applied in the Muon Ionization Cooling Experiment (MICE), are critical for the structure design and mechanical stability because of a large diameter and relative high magnetic field. This paper presents an analytical stress solution for the MICE coupling coil assembly. The stress due to winding tension is calculated by assuming the coil package as a set of combined cylinders.The thermal and electromechanical stresses are obtained by solving the partial differential equations of displacement based on the power series expansion method. The analytical stress solution is proved to be feasible by calculating stresses in a tested superconducting solenoid with 2.58 m bore at room temperature. The analytical result of the MICE coupling coil is in good agreement with that ofthe finite element which shows that the transverse shear stress induced by Lorentz force is principally dominant to magnet instability.     

2D thermal analysis for heat transfer from casing to winding pack in JT-60SA TF coils

In the framework of the Broader Approach Agreement, Europe is involved in the design activities for the Japanese Tokamak JT-60SA, investigating, among several issues, the operation of the superconducting TF magnets and their subsystems, aimed at the reactor conceptual design definition. In particular, one of the main critical aspects to study is the heating of the conductor due to both direct component of energy deposited by neutrons and by secondary gamma generated during plasma operation, and heat generated by the radiation on casing and transferred to the winding pack. Indeed, the operating temperature and the relevant temperature margin (i.e. the operating safety margin) of the magnet will depend strongly on the heat loads and on the capability of the coolant to remove it. Furthermore, the heat power to the conductor will depend on several aspects, namely the thickness of insulating material, the mass flow rate of helium flowing in the conductors and its thermodynamic properties at operating conditions, and the layout of the superconductors constituting the winding. Moreover, a crucial aspect in the final design will be the presence and position of the casing cooling channels. In this paper a 2D sensitivity analysis of heat transfer from casing to winding pack with respect to cooling channels number and position is presented, based on the reference layout of the magnet. As a result, we evaluated the optimum number and positioning of cooling channels needed, as a trade-off between magnet operating limits and available cryogenic power and if, at limit, they could be even neglected in normal operation, keeping dwell-time within reasonable values.     

High current superconductors for DEMO

In the assumption that DEMO will be an inductively driven tokamak, the number of load cycles will be in the range of several hundred thousands. The requirements for a new generation of Nb₃Sn based high current conductors for DEMO are drafted starting from the output of system code PROCESS. The key objectives include the stability of the DC performance over the lifetime of the machine and the effective use of the Nb₃Sn strand properties, for cost and reliability reasons. A preliminary layout of the winding pack and conductors for the toroidal field magnets is presented. To suppress the mechanism of reversible and irreversible degradation, i.e. to preserve in the cabled conductor the high critical current density of the strand, the thermal strain must be insignificant and no space for micro-bending under transverse load must be left in the strand bundle. The “react-and-wind” method is preferred here, with a graded, layer wound magnet, containing both Nb₃Sn and NbTi layers. The implications of the conductor choice on the coil design and technology are highlighted. A roadmap is sketched for the development of a full size prototype conductor sample and demonstration of the key technologies.     

Process Design of Cryogenic Distribution System for CFETR CS Model Coil

The superconducting magnet of Central Solenoid(CS) model coil of China Fusion Engineering Test Reactor(CFETR) is made of Nb_3Sn/Nb Ti cable-in-conduit conductor(CICC),and operated by forced-flow cooling with a large amount of supercritical helium.The cryogenic circulation pump is analyzed and considered to be effective in achieving the supercritical helium(SHe) circulation for the forced-flow cooled(FFC) CICC magnet.A distributed system will be constructed for cooling the CFETR CS model coil.This paper presents the design of FFC process for the CFETR CS model coil.The equipment configuration,quench protection in the magnet and the process control are presented.     

CEA studies and qualifications prior to the JT-60SA TF coils manufacture

Following the first conceptual design activity in which the general design of the JT-60SA TF system was defined in agreement with all the participants in the project (CEA, ENEA, F4E), a second phase dealing with the detailed design was engaged by each of the voluntary contributors. For the French part which includes the procurement of 9 of the TF winding packs and their integration in the casing, an industrial contract was signed mid 2011 with Alstom (France). Several actions have been carried out to prepare the manufacturing phase.To precisely define one of the main interfaces which is the temporary electrical connection of the coils to the current leads during cold test in the CEA facility as well as their final connection to the feeders at the Naka site, a design compatible with both requirements was developed by CEA, supported by the previous developments led on the joints and assembly techniques.In addition to prepare the coils manufacture, hydraulic qualification was led on the first conductor qualification length to set the parameters which will be used by the coils manufacturer for conductor acceptance.At last, mechanical characterizations of both the conductor and of the empty compacted jacket were performed in order to define as precisely as possible the elastic and plastic properties of these components. These are crucial properties used during the bending process which is one of the most important operations during the winding pack manufacture. These data will be very helpful for the winding machine parameters settings as well as for designing the local bending tooling needed to shape the conductors extremities at the connection area and at the double pancakes joggles.     

Magnet design of toroidal field coils for the UWMAK-Tokamak systems

The design of toroidal field coils for the UWMAK series of Tokamak reactor designs is described. These are cryogenically stable coils cooled in liquid helium to 4.2 K. Each individual turn of composite conductor of TiNb plus matrix conductor is epoxied into a groove in a thin disk structure. The magnet is divided into 12, 18 or 24 sectors; each sector is comprised of 15–20 thin disks which are spaced and bolted together to form a rigid structure with all disk surfaces exposed to cooling. The overall shape of each ‘D’ magnet sector is chosen so that only constant tension forces are present. Bending forces do occur but only near transition sections from the D to the central straight section of each coil. This method of rigid mounting should be compared with loose ‘jelly-roll’ windings on a central coil form, a more typical magnet fabrication technique. The design procedure is for the composite conductor TiNb plus copper (or aluminum) to be mounted in stainless steel (or aluminum alloy) disks. Full stability is obtained for strains less than 0.2% for steel support and less than 0.4% for aluminum supports based on stress-strain resistivity experiments in progress. The use of high purity aluminum conductor and high strength aluminum alloy structure reduces costs significantly dependent only on the orderly development of new aluminum TiNb composite conductors. The amount of TiNb is conservatively chosen to carry full current at 5.2 K although operation at 4.2 K is planned and full recovery to the superconducting state could be obtained with full current wire quantities selected at 4.3 K. This conservative choice doubles the amount of TiNb used at 8 tesla but provides an extra temperature rise of ΔT = 0.9 K above expected usual temperature excursions.Magnet safety and protection is based on the natural mutual coupling of many coils which are closely coupled to each other. If one coil loses current, the other coils increase their currents to keep the flux as constant as possible. The uncoupled flux and companion field energy would be discharged by a high voltage power supply temporarily set to discharge the one bad coil. Such sub-division and partial energy removal requires that there be substantial subdivision of coils into many separate dewars, so that problems can be isolated. An expression for the magnetic forces on sectioned toroidal field coils is given in closed form and is used to compute the shape of a specific coil. Data obtained here are shown to be in good agreement with those given by more complex procedures. The most severe structural design requirement is based on simultaneous loss of current in two adjacent sectors. The remaining sectors attempt to straighten out into a solenoid which compresses the structure between coils except beside the bad coil or coils where tension might exist. Such current loss in two adjacent sectors is considered an extremely unlikely occurrence since the discharge procedure mentioned above takes place in less than 1 min so that simultaneous refers to a 1 min overlap. Because of such rapid amelioration of the causes of current change and flux motion, no temperatures can exceed room temperature during the orderly shutdown of one or two coils. In general, the study illustrates that fully stable magnets using composite conductors should be engineered without major uncertainties according to straightforward scientific concepts. While subsequent designs will undoubtedly include improvements there is no reason to expect that superconductivity implies venturesome unknown TF coil performance.     

Quench analysis of an ITER TF coil

The first realistic application of the recently developed 4C code is presented, aimed at showing its capability to simulate in an integrated fashion relevant transients for the superconducting coil operation in the International Thermonuclear Experimental Reactor (ITER), both at the system and at the conductor levels. The quench initiation and propagation in an ITER TF coil is considered, including the coil fast discharge phase until the opening of the relief valves. The 14 coil pancakes cooled by alternating clockwise/counter-clockwise SHe flow, the radial plates and the case, the 96 case cooling channels, and the external cryogenic circuit up to the LHe bath are included in the model. The results of the analysis are discussed with particular reference to quench propagation in the winding, hot spot temperature and peak pressurization, mass flow rate evolution in the different system components. The accuracy of the analysis is guaranteed by suitable convergence studies.     

Stress and Thermal Analysis of the In-Vessel RMP Coils in HL-2M

A set of in-vessel resonant magnetic perturbation(RMP) coils for MHD instability suppression is proposed for the design of a HL-2M tokamak.Each coil is to be fed with a current of up to 5 kA,operated in a frequency range from DC to about 1 kHz.Stainless steel(SS) jacketed mineral insulated cables are proposed for the conductor of the coils.In-vessel coils must withstand large electromagnetic(EM) and thermal loads.The support,insulation and vacuum sealing in a very limited space are crucial issues for engineering design.Hence finite element calculations are performed to verify the design,optimize the support by minimizing stress caused by EM forces on the coil conductors and work out the temperature rise occurring on the coil in diferent working conditions,the corresponding thermal stress caused by the thermal expansion of materials is evaluated to be allowable.The techniques to develop the in-vessel RMP coils,such as support,insulation and cooling,are discussed.     

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.     

Fabrication and installation of equilibrium field coils for the JT-60SA

Recently, fabrication of the first superconducting coil in JT-60SA tokamak (EF4 coil) was finished. EF4 coil has ten double pancake (DP) coils. All DP coils were stacked up to form the winding pack. In order to check the manufacturing error of DP coils, their circularities (in-plane ellipticities) were evaluated for all DP coils. Positions of conductors for each DP coil were measured before curing process. Error bars of the current centers, which were used for the index for DP coil's circularity, ranged between 1.1 and 2.5 mm. During stacking the DP coils, the positions of these coils were optimized in order to cancel the error of circulation of the winding pack. As the result, the manufacturing error of the radial current center was achieved 0.6 mm for the winding pack. This value was an order of magnitude smaller than the required manufacturing error of EF4 coil.