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。     

ITER纵场磁体过渡馈线结构力学分析

运用数值方法计算了不同等离子体运行时刻纵场磁体过渡馈线(CFT)超导母线上的电磁载荷,并确定了磁感应强度最大的时刻,采用增量有限元法对过渡馈线进行非线性力学分析,得到不同工况下结构上的应力分布及变形情况。分析结果表明,带有万向节的过渡馈线结构具有足够的强度来承受运行过程中的各种载荷,从而证明了结构设计的合理性。     

北京正负电子对撞机重大改造工程中超导四极磁体传热与流动计算

超导四极(SCQ)磁体是北京正负电子对撞机重大改造工程(BEPCⅡ)的关键设备之一。本文对SCQ磁体恒温器进行稳定运行状态下传热和流动计算。计算得到了磁体在低温下的热负荷以及磁体恒温器内各组成部分的温度分布，并在此基础上，提出减小SCQ磁体热负荷的方法。比较计算了SCQ磁体采用超临界和过冷液氦两种冷却方式对磁体稳定运行的影响。     

多节热离子能量转换器的节间换向连接件热应力仿真设计

正节间换向连接件(简称连接件)是一种用于多节热离子能量转换器的补偿性元件,依据其功能特性提出了开槽缸结构设计。通过运用有限元仿真软件,建立起仿真体模型,参数化扫描后得到多组设计组合数据的热应力结果,最终依据热应力结果得到连接件设计结构的最优几何参数。最后针对应力集中问题,对槽末端进行了开孔设计,最大应力得到了降低。     

国际热核聚变实验堆磁体线圈终端盒的设计

线圈终端盒（CTB）是国际热核聚变实验堆超导磁体系统的重要组成部分，其内部组件的漏热常常是整个磁体系统的主要漏热源之一，在很大程度上决定着低温系统的液氦消耗量。本文从降低热负荷的角度对CTB内部冷屏、超导电流传输线、电流引线、阀及冷却管路、外部盒体的设计进行了详细阐述，为最终结构的确定提供了理论依据。     

反应堆吊篮的非轴对称温度场和热应力分析

反应堆吊篮在运行时受到γ射线辐射后会引起内部发热,由于燃料组件在堆内排列是非轴对称的,因此在吊篮内壁所产生的内热源也是非轴对称的。本文推导了由非轴对称内热源引起的温度场求解方程,并用热弹性位移势与应力函数相融合方法求解非轴对称的热应力解析式。以30万千瓦秦山核电站所设计的反应堆吊篮为例,计算了稳态温度场和热应力。计算结果表明温度应力环向分布不均匀发比较严重,如把问题简化为轴对称处理,则计算所得的热应力作为设计依据是不恰当的。     

设计HLS储存环高频腔的有限元分析

描述了对ＨＬＳ储存环设计高频腔进行有限元结构分析的过程和得到的结果,诸如位移,应力,温度场,热应力,热变形,以及机械固有频率和主振型先进。这些量对高频腔水冷和结构设计有十分重要的意义。     

北京谱仪Ⅲ束流铍管环形窄通道冷却优化设计

北京谱仪Ⅲ(BESⅢ)要求束流铍管部分物质量尽可能少,同时要求将束流铍管内部产生的热负荷通过冷却介质带走,精确控制外壁温度。本文结合理论分析和有限元数值模拟对束流铍管的冷却环形窄通道尺寸进行优化设计,确定了窄通道最佳间隙,研究了集中热负荷对内外壁温度场的影响,并确定了冷却液最佳工作流速。根据设计结构,制作出束流铍管实际尺寸实验模型。实验测量结果表明,束流铍管冷却结构设计合理,最佳流速满足冷却要求。     

高通量工程试验堆燃料元件动态热工参数测量——全厂断电、停堆冷却与剩余发热实验

作者应用高通量工程试验堆堆芯燃料元件温度-流量测量装置测定了在全厂断电事故情况下的燃料元件热工参数(元件盒进出口水温,元件包壳温度,元件盒流量及其热功率)的瞬态过程,测定了在停堆冷却过程中启停事故泵时的流动反向过程,进行了停堆后的长时间剩余发热测量,给出了上述测量结果。     

冷冻靶制备中温度控制数值模拟

在二维轴对称模型下，以及惯性约束核聚变冷冻靶制备的温度控制过程中，利用计算流体力学程序Fluent，对聚变腔内的温度场变化进行模拟。研究了腔内气体的自然对流效应对冷冻靶温度分布的影响，模拟了通过在冷却环上施加一正弦振荡的温度场来降低冷冻靶内表面粗糙度的过程，给出了动态快速冷冻方法中的靶温度随冷却环温度的变化过程。     

ITER重力支撑的制造设计、认证测试及关键技术研究

重力支撑（GS）作为国际热核聚变实验堆（ITER）磁体支撑系统的关键部件,不但要承受环向场超导磁体净重以及交变的电磁载荷,同时还需隔离来自杜瓦环的热量以维持环向场超导线圈的热稳定性。本文通过有限元分析和工程测试验证了GS结构设计的可靠性;通过换热分析和真空热交换效率测试验证了热锚连接结构的可靠性;通过全尺寸螺栓77 K疲劳测试验证了螺栓原型件的疲劳性能。在随后的制造过程中,使用液压拉伸器和研制的高精度螺栓伸长量测量装置对所有的螺栓进行了均匀、精确地紧固。真空正压氦检漏的测试结果证明了GS的泄漏率远低于ITER的要求。基于以上工程测试的结果,本文设计的GS的结构是可行的且能运用于ITER装置中。      

Thermal analysis of the cryostat feed through for the ITER Tokamak TF feeder

In Tokamaks,the toroidal field (TF) coil feeder is an important component that is used to supply the cryogens and electrical power for the TF coils.As a part of the TF feeder,the cryostat-feed through (CFT) is subject to low temperatures of 9 and 80 K inside and room temperature of 300 K outside.Based on the features of the International Thermonuclear Experimental Reactor TF feeder,the thermal performance of the CFT under the nominal conditions is studied.Taking into account the conductive,convective and radiation heat transfer,the finite element model of the CFT is built.Transient thermal analysis is performed to determine the temperatures of the CFT on the 9th day of cooldown.The model is assessed by comparing the cooling curves of the CFT after 9 days.If the simulation and experimental results are the same,the finite element model can be considered as calibrated.The model predicts that the cooling time will be approximately 26 days and the temperature distribution and heat load of the main components are obtained when the CFT reaches thermal equilibrium.This study provides a valid quantitative characterization of the CFT design.     

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.     

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.     

Transient electrical behaviour of the ITER TF coils during fast discharge and two fault cases

Improvements of high voltage design criteria and quality assurance for ITER coils are indispensable taking into account the problems occurred during high voltage tests of the ITER TF model coil. One important aspect to consider is the transient electrical behaviour because fast changes of voltages may cause local overloading and destruction of the insulation system. This paper will present the calculation of the terminal voltages within the ITER TF coil system and the voltage stress of the insulation within an individual ITER TF coil for the fast discharge and two fault cases. Proposals for the high voltage tests are discussed based on the calculated voltage stress of the two fault cases and the experiences gained during the ITER TF Model Coil test to ensure appropriate dielectric quality of the ITER TF coils.     

Decay Heat Removal by Natural Circulation of Vacuum Vessel Coolant for ITER

The decay heat-driven temperature transients of the in-vessel components following a postulated loss of all in-vessel cooling have been calculated. The resulting time-dependent heat load to the vacuum vessel is due to radiation from the backplate and convection of postulated steam between backplate and vacuum vessel. It is shown, that even for a failure of all in-vessel cooling and total loss of power, the ITER design can rely on passive decay heat removal by natural circulation in one of the two existing cooling loops of the vacuum vessel. A mathematical model describes the transient operating conditions and shows that the temperature established by natural circulation does not exceed 200°C at the maximum shut down heat load to the vacuum vessel. Therefore, no additional emergency cooling system is required if the existing heat exchanger is designed for natural circulation and a bypass is used during normal operation to maintain operation temperature.     

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.     

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.     

Design of Tokamak ELM Coil Support in High Nuclear Heat Environment

In Tokomak, the support of the ELM coil, which is close to the plasma and subject to high radiation level, high temperature and high magnetic field, is used to transport and bear the thermal load due to thermal expansion and the alternating electromagnetic force generated by high magnetic field and AC current in the coil. According to the feature of ITER ELM coil, the mechanical performance of rigid and flexible supports under different high nuclear heat levels is studied. Results show that flexible supports have more excellent performance in high nuclear heat condition than rigid supports. Concerning thermal and electromagnetic （EM） loads, optimized results further prove that flexible supports have better mechanical performance than rigid ones. Through these studies, reasonable support design can be provided for the ELM coils or similar coils in Tokamak based on the nuclear heat level.     

Cyclic testing of shear keys for the ITER magnet system

Shear keys are to be used to support the out-of-plane loading of the toroidal field (TF) coils during a plasma pulse in ITER. At the inner intercoil structures (IIS) a set of poloidal shear keys is used to take the shear load at each connection between adjacent TF coils. Solid circular keys have been selected as reference. At the outer intercoil structures (OIS) adjustable conical shear keys and friction joint based shear panels are used to take the shear load. Low voltage electrical insulation is required at the flanges of the IIS and OIS, plus for all the bolts, poloidal keys and adjustable keys. This electrical insulation has to withstand large compression associated with some shear or slippage. A ceramic coating was selected for this purpose. The main scope of the experimental campaign was the mechanical testing of the shear keys and the electrical insulation in operational conditions relevant to ITER. Both keys were made of Inconel 718, provided with a ceramic alumina coating and inserted into flanges made of cast AISI 316 LN. The adjustable conical shear key was pre-loaded at room temperature and subject to cyclic shear loads of 2.5 MN for a large number of cycles (about 30,000) at cryogenic temperature (77 K). The conical key and the alumina coating remained undamaged after the test. Another test campaign was then performed with higher shear loads (up to 3 MN) to reach a sufficient safety margin even with the friction effect due to the pre-load. A set of 15,000 cycles were completed followed by some cycles at higher loads to reach the ultimate limit, which is the shear load to be experienced by the key in case of a poloidal field (PF) coil short.