SST-1托卡马克绝缘子优化设计

对SST-1托卡马克重要部件复合材料绝缘子进行有限元分析,获得绝缘子的高压绝缘性能及低温力学性能,并根据分析结果,对结构进行优化。结果显示,优化后结构提高了绝缘子的绝缘性能及低温力学性能,可满足SST-1托卡马克对绝缘子的设计要求。     

高温气冷堆氦气环境中电气设备绝缘设计研究

应用巴申定律研究了氦气的电气击穿特性,并与空气的绝缘特性进行比较。以高温气冷堆氦气透平发电系统电机腔室的设计参数为例,结合氦气的巴申曲线,对氦气条件下气体压力和极间距离的关系进行深入探讨,并提出氦气环境中电气设备绝缘设计需关注的问题。研究结果表明,氦气最小击穿电压为150~200V,绝缘特性较差,电气设备绝缘结构设计应考虑氦气环境压力的影响,现有针对压水堆电站电气设备绝缘结构的验收准则和试验方法并不完全适用于氦气环境。     

几种绝缘薄膜开关特性之比较

介绍了国内外几种绝缘薄膜开关 ,从开关的结构、对应的输出方波波形和输出阻抗等几个方面进行了比较 ,分析了其各自的优缺点     

管道密封绝缘接头的非线性分析

本文所分析的管道密封绝缘接头是一种特殊的管道接头，广泛应用于各种工业领域。本文使用非线性有限元程序MSC．MARC，采用弹性应变能理论和自动接触分析等技术，对该管道密封绝缘接头进行了接触状态下的超弹性分析，分析密封绝缘接头预紧后在10MPa压力作用下的变形、应变和应力分布、接触力。给出了对结构改进的建议，说明了MSC．MARC非线性有限元软件在核工程设计研究中有广阔的应用前景。     

SOI材料和器件及其应用的新进展

综述了绝缘层上的硅(SOI)材料的新结构包括不同绝缘埋层和不同半导体材料结构的最新进展，介绍了SOI器件的新结构和SOI器件在抗辐射电子学方面的应用，报道了国内在SOI技术的研发和产业化的最新动态。     

核电厂安全壳中压电气贯穿件绝缘支撑盘绝缘耐潮性能试验分析

针对中国核动力研究设计院（NPIC）设计生产的中压电气贯穿件的绝缘支撑盘注塑材料聚砜,测试其吸水率、介质损耗、相对介电常数,绝缘电阻、体积电阻率、表面电阻率、介电强度等随环境温度和相对湿度的变化趋势。测试结果表明,环境相对湿度对聚砜材料的吸水率影响显著,23℃时湿度由30%增至98%,对应吸水率由0.012%增至0.106%,增幅高达783.3%;绝缘电阻随环境温度和相对湿度的增大而逐渐减小,降幅最高达99.82%,但绝缘支撑盘相间绝缘电阻始终大于200 TΩ;绝缘电阻受表面电阻率影响较大,绝缘电阻和表面电阻率随环境温度和相对湿度的变化趋势非常接近;相对介电常数和介电强度受环境温湿度影响很小。因此,NPIC设计生产的中压电气贯穿件绝缘支撑盘耐潮性能和电气绝缘性能优良,能够在高电压和高温高湿环境中稳定可靠工作。      

200千伏直流绝缘变压器的试制

在高压倍加器中,通常采用带有绝缘传动件(绝缘带或绝缘轴)的电动发电机组或绝缘变压器作为高压整流管灯丝加热的供电电源。电动发电机组虽然比较经济,但存在以下的缺点:(1)直流发电机工作不可靠,可能使整流管熄灭,从而引起整流管的击穿;(2)绝缘传动件容易     

静电场分析法优化设计负离子源中性束注入系统高压仓结构

为探究最为合理的负离子源中性束注入(NNBI)装置高压仓(HVD)结构,采用有限元静电场分析方法,对不同结构高压仓进行仿真计算分析,在考虑分布电容影响的前提下,优化设计了高压仓的主体和外形结构,并确定了安全绝缘距离,对高压仓的工程研制提供了理论指导,为NNBI的安全稳定运行奠定了基础。     

CPR1000核电厂反应堆冷却剂泵电机绝缘问题分析

文章介绍了CPR1000反应堆冷却剂泵（主泵）电机轴电压产生的原理,针对轴电压对主泵电机设备运行产生的危害和影响,剖析影响主泵电机轴绝缘失效的原因,通过主泵电机轴绝缘故障问题实例,阐述了轴绝缘故障排除处理方法,并提出了几种改善轴绝缘的相关措施。     

基于绝缘芯平面变压器的新型高压直流电源的调试及输出特性

介绍了一种以绝缘芯平面变压器为核心部件的新型高压直流电源的原理和基本结构,并给出了自行设计的实验样机的参数。对样机进行了调试,并测量了不同负载下电源的输出特性,发现在负载较重时电源会出现明显的压降。建立合理的等效电路和参数模型,对导致压降的原因进行了理论分析,进而提出了相应的解决方法,并进行了实验验证。结果证明,在负载较重时,绝缘芯平面变压器中的压降主要由漏磁引起,此问题通过调整功率因数解决。     

Cryogenic mechanical testing of ITER prototype axial breaks

Within the superconducting magnet program for ITER, cryogenic components need to be tested to verify their design. Especially for the helium supply of the magnet system, feeders are needed integrating at the same time high voltage insulation to separate the inner magnet system electrically from the outer cryostat shell. Beside of high voltage and helium flow properties, these axial breaks will be exposed to a limited mechanical loading during operation of the magnet system. Therefore, mechanical tests needs to be performed at room temperature as well as at cryogenic temperature of 77 K.A possible breaker design was provided by Babcock Noell. To verify this design mechanically quasi-static and fatigue tests under bending, torsion and axial loading were done. Results on the performance of the prototypes are presented approving a superior mechanical quality.     

High Temperature Baking Test and Study on Axial Insulation Breaks

Axial insulation breaks are needed in ITER superconducting magnet system, which are used for separation of high voltage area from grounded cooling pipes system. To determine the maximum safe temperature that the insulating break can withstand without damage, such as preventing damage by overheating during welding of the insulating breaks to the helium cooling pipes for the superconducting magnets and high temperature baking to remove moisture, the glass transition temperature was tested by using the dynamic mechanical analysis with the standard sample made from epoxy resin. Furthermore, the high temperature baking test of axial insulation break was performed, further helium tight test and high voltage tests indicate the baked insulation break is in good condition.     

Non-superconducting magnet structures for near-term, large fusion experimental devices

Water cooled copper magnets provide a means of producing high magnetic fields for tokamaks using a well developed existing technology. The basic function of these magnets is to provide reliable, both time varying and steady state, magnetic fields. Copper electrical properties, insulation, and water cooling systems play major roles in design selection. Aside from being electro-magnetic devices, coils designed for tokamaks must be self-supporting structures, capable of resisting large I × B magnetic forces. These magnets require the integration of both electrical and structural design considerations.Magnet integrity is enhanced by the presence of structures which lend additional external support. These external structural systems are highly stressed and, often, deflection limited.This paper describes the magnet and structural design in the following American tokamak devices: the Princeton Large Torus (PLT), the Princeton Divertor Experiment (PDX), and the Tokamak Fusion Test Reactor (TFTR). The Joint European Torus (JET), also presented herein, has a magnet structure evolved from several European programs and, like TFTR, represents state of the art magnet and structure design.The PLT device was designed in 1971 as a high plasma current tokamak. At the time it incorporated the latest in copper magnet and structure technology. Design features on this machine have in some fashion subsequently been incorporated on every major device built within the tokamak fusion community.     

Thermal and structural analysis of the W7-X magnet heat radiation shield

The thermal insulation of the W7-X – cryostat consists of multi-layer insulation (MLI) and an actively cooled thermal shield. The shield is cooled by He gas flowing in pipes, which are flexibly attached via copper braids. The paper presents the basic mechanical and thermal layout of the complex plasma vessel shield, which is made of a glass fibre compound with three embedded copper nets.Main mechanical loads on the shield are electromagnetic forces resulting from rapid shut down of the magnet system, and the self weight. Design and calculations were performed iteratively to determine the number as well as orientation of electrical insulation gaps within the copper nets, and the number and positions of the supports.It is shown that the maximum displacements of the panels, the maximum forces on the supports, and the shield temperatures fulfil the requirements.     

Design and Fabrication of a Hybrid HTS Magnet for 150 kJ SMES

This paper describes design and fabrication of a hybrid high temperature superconducting (HTS) magnet for a 150 kJ superconducting magnetic energy system. The hybrid HTS magnet, which employs both BSCCO taps and YBCO taps, is composed of 18 double pancake coils (DPC). Six DPC made of BSCCO are respectively installed on the top and bottom with an inner diameter of 240 mm, an outer diameter of 396 mm, and height of 11.5 mm. Six YBCO coils are mounted in the middle of the magnet with the inner and outer diameters of 264, 396 mm and height of 12.1 mm. Copper plates with thickness of 1 mm are arranged between DPC for the cooling of the heat, and epoxy resin plates with thickness of 0.2 mm are arranged between coil and copper plate for the insulation. The inductance of the magnet is about 11.06 H and the resistance is 169.5 Ω at the room temperature. In order to evaluation the performance of the hybrid magnet, each solidified DPC and the assembled magnet were tested in a bath of liquid nitrogen at 77 K. The resistances of 17 joints were also measured and evaluated by the standard four-probe method.     

Preliminary design of KTX magnet system

KTX is a reversed field pinch magnetic confinement device of which the magnet system is designed in ASIPP and USTC. The main parameter of KTX is between RFX and MST. Its magnet system includes the toroidal field (TF) winding and poloidal field (PF) winding (ohmic heating winding and equilibrium field winding), which are less complex than tokamak device due to the fact that a tokamak requires a superconducting system to perform quasi-steady state operation and achieve Q > 10. However, the most important part of the magnet system design lies in how to keep the TF magnetic field ripple, as well as any kinds of stray field, to a minimum value. The main design activities of the KTX magnet system are presented as detailed as possible in this paper, and the main activities which have already been completed include magnet coils position and winding, insulation design, plasma modeling prediction, thermal analysis, magnetic field calculations were analyzed and so on. The magnet system design is one of the major activities for KTX device design, which is effective guarantee for the future R&D and manufacture. Besides, the detailed design activities should be continuously optimized and changed based on the results from future R&D and relevant tests.     

Magnet Safety Assessment for ITER

The ITER magnet system consists of structurally linked sets of toroidal (TF) and poloidal (PF) field coils, central solenoid (CS), and various support structures. The coils are superconducting, force flow Helium cooled with a Kapton-Glass-Epoxy multilayer insulation system. The stored magnetic energy is about 100GJ in the TF system and 20GJ in the PF-CS. Coils and structure are maintained at 4 K by enclosing them in a vacuum cryostat. The cryostat, comprising an outer envelope to the magnets, forms most of the second radioactivity confinement barrier. The inner primary barrier is formed by the vacuum vessel, its ports and their extensions. To keep the machine size within acceptable bounds, it is essential that the magnets are in close proximity to both of the nuclear confinement barriers. The objective of the magnet design is that, although local damage to one of the barriers may occur in very exceptional circumstances, large scale magnet structural or thermal failure leading to simultaneous breaching of both barriers is not credible. Magnet accidents fall into three categories: thermal (which includes arcing arising from insulation failure and local overheating due to discharge failure in the event of a superconductor quench), structural (which includes component mechanical failure arising from material inadequacies, design errors and exceptional force patterns arising from coil shorts or control failures), and fluid (Helium release due to cooling line failure). After a preliminary survey to select initial faults conceivable within the present design, these faults are systematically analyzed to provide an assessment of the damage potential. The results of this damage assessment together with an assessment of the reliability of the monitoring and protective systems, shows that the magnets can operate with the required safety condition.     

Modeling of a confinement bypass accident with CONSEN,a fast-running code for safety analyses in fusion reactors

The CONSEN (CONServation of ENergy) code is a fast running code to simulate thermal-hydraulic transients, specifically developed for fusion reactors. In order to demonstrate CONSEN capabilities, the paper deals with the accident analysis of the magnet induced confinement bypass for ITER design 1996. During a plasma pulse, a poloidal field magnet experiences an over-voltage condition or an electrical insulation fault that results in two intense electrical arcs. It is assumed that this event produces two one square meters ruptures, resulting in a pathway that connects the interior of the vacuum vessel to the cryostat air space room. The rupture results also in a break of a single cooling channel within the wall of the vacuum vessel and a breach of the magnet cooling line, causing the blow down of a steam/water mixture in the vacuum vessel and in the cryostat and the release of 4 K helium into the cryostat. In the meantime, all the magnet coils are discharged through the magnet protection system actuation. This postulated event creates the simultaneous failure of two radioactive confinement barrier and it envelopes all type of smaller LOCAs into the cryostat. Ice formation on the cryogenic walls is also involved. The accident has been simulated with the CONSEN code up to 32 h. The accident evolution and the phenomena involved are discussed in the paper and the results are compared with available results obtained using the MELCOR code.     

Manufacturing and assembly of the plasma- and outer vessel of the cryostat for Wendelstein 7-X

Wendelstein 7-X is an advanced helical stellarator, which is presently under construction at the Greifswald branch of IPP. A set of 70 superconducting coils arranged in five modules provides a twisted shaped magnetic cage for the plasma and allows steady state operation. Operation of the magnet system at cryogenic temperatures requires a cryostat which provides thermal protection and gives access to the plasma. The main components of the cryostat are the plasma vessel, the outer vessel, the ports, and the thermal insulation. The German company, MAN Diesel & Turbo SE Deggendorf (former MAN DWE GmbH Deggendorf), is responsible for the manufacture and assembly of the plasma vessel, the outer vessel and the thermal insulation. This paper describes the manufacturing and assembly technology of the plasma and outer vessel of the cryostat for Wendelstein 7-X.     

Neutron irradiation effects on superconducting wires and insulating materials

On the progress of the Deuterium–Deuterium (D–D) or Deuterium–Tritium (D–T) burning plasma devices, the importance of neutron irradiation on superconducting magnet materials increases and the data base is desired to design the next generation devices. To carry out the investigations on the effect of neutron irradiation, neutron irradiation fields are required together with post-irradiation test facilities. In these several years, a collaboration network of neutron irradiation effect on superconducting magnet materials has been constructed. 14 MeV neutron irradiation was carried out at Fusion Neutronics Sources (FNS) in Japan Atomic Energy Agency (JAEA) and fission neutron irradiation was performed at JRR-3 in JAEA. After the irradiation, the Nb₃Sn, NbTi and Nb₃Al samples were sent to High Field Laboratory for Superconducting Materials (HFLSM) in Tohoku University and the superconducting properties were evaluated with 28 T hybrid magnet. Also, the organic insulation materials are considered to be weaker than superconducting materials against neutron irradiation and cyanate ester resin composite was fabricated and tested at the fission reactor. One clear result on Nb₃Sn was the property change of Nb₃Sn by 14 MeV neutron irradiation over 13 T. The critical current was increased by 1.4 times around 13 T but the increment of the critical current became almost zero at higher magnetic fields and the critical magnetic field of the irradiated sample showed almost the same as non-irradiated one.