基于CODAC的ITER电流引线测控系统

中国科学院等离子体物理研究所(Institute of Plasma Physical,Chinese Academy of Sciences,ASIPP)负责国际热核聚变实验堆(International Thermonuclear Experimental,ITER)60根高温超导电流引线(High Temperature Superconducting Current Lead,HTSCL)产品的研制与测试,并在2013–2015年间开展了三对三种电流等级(68 k A、55 k A和10 k A)的高温超导电流引线认证制造。为检验电流引线的低温大电流性能,ASIPP与印度塔塔咨询服务公司(Tata Consultancy Service,TCS)及ITER合作开发了基于CODAC(Control,Data Access and Communication)框架的ITER高温超导电流引线测控系统。该系统包括传统西门子PLC300工艺过程测控系统、基于Lab VIEW的失超保护系统、基于PLC400冗余设计的互锁系统和基于NI c系列模块的快速控制系统(Plant system Controller,Fast Controls,PCF)。目前本系统已通过三轮验收测试并在2015年1月份的ITER CC 10 k A电流引线原型件和同年7月份的ITER TF 68 k A电流引线原型件中成功应用。结果表明,本系统能很好地满足电流引线的实验需求,得到ITER国际认同。电流引线测控系统软硬件遵照ITER的CODAC标准进行设计,是CODAC和互锁保护规范的首次在ITER真实组件物理性能测试的联合应用案例,可作为ITER采购包出厂验收推行的CODAC模范。     

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。     

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。     

聚变原型超导线圈试验成功

【美国《科学日报》2008年9月14日报道】在欧洲委员会（EC）、日本原子能研究开发机构（JAEA）和国际热核聚变实验堆（ITER）组织的支持下,欧洲的聚变研究机构“用于能源的聚变”（Fusion for Energy）已成功对铌．钛制成的ITER极向励磁线圈原型超导体进行了试验,该超导体在6．4T磁场和52kA电流下达到稳定运行。极向励磁线圈将用于维持ITER装置内部的等离子体平衡和位形。     

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。     

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。     

超导磁体气冷电流引线的优化设计

从超导磁体气冷电流引线的经典微分方程出发,将电流引线分为很少的几段,提出了一种较为精确计算电流引线长横比及由电流引线末端流入低温容器热量的计算方法。并以黄铜为例计算了电流引线的长横比和流入低温容器的热量。     

SSRF储存环注入脉冲电源的设计

本文介绍了采用固态半导体放电开关实现的1.7 kV、10 kA的Septum电源和7.8 kV、3.7 kA的Kicker电源,详细分析了两台脉冲电源的电容器充电电源、脉冲放电开关、放电回路结构、热漂移校正等问题.原型样机的结果表明,其电流幅度稳定性都优于±0.1%,其他主要指标均可达到设计要求.     

国际热核聚变实验堆标准体系研究

国际热核聚变实验堆(ITER)是目前世界上最大的磁约束核聚变试验堆。文章按照国家标准GB/T 13016—2018《标准体系构建原则和要求》中规定的方法,对ITER标准体系进行了研究,给出了构建ITER标准体系的原则、构建ITER标准体系的方法和步骤以及ITER标准体系结构图,介绍了ITER标准体系标准明细表的总体情况,为我国开展类似核聚变反应堆设计和建造的提供重要参考。     

ITER屏蔽包层活化分析

作为国际热核聚变实验堆(ITER)的重要部件之一,屏蔽包层承受高强度聚变中子辐照,需要定期更换和维修。当活化的屏蔽包层从ITER托卡马克装置移到热室时,可能会给工作人员造成严重的辐射照射,是ITER大厅和热室屏蔽设计的重要辐射源。文中基于ITER最新中子学分析基准模型和"二步法"停堆剂量计算方法,使用超级蒙特卡罗核计算仿真软件系统SuperMC针对15号屏蔽包层建立精细的中子学模型,并计算分析包层的活化情况及最严重情况下的周围辐射剂量率,并初步应用于ITER赤道窗口室的屏蔽分析。计算结果显示,单个包层周围最大剂量率为350 Sv/hr,当传送小车停留在赤道窗口室内时,窗口室屏蔽门外剂量率高于10 mSv/hr,不足以满足设计要求。     

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.     

High temperature superconductors for the ITER magnet system and beyond

The use of high temperature superconductor (HTS) materials in future fusion machines could increase the efficiency drastically, but strong boundary conditions exist. To outline the prospects, challenges and problems, first the benefit of using HTS materials is estimated considering the saving in cryogenic power. Next, it is demonstrated that industrial available HTS materials can be used for fusion today. For this purpose, we give a short summary of results that have been obtained from an ITER conform 70 kA HTS current lead that was designed, built and tested by the Forschungszentrum Karlsruhe and the CRPP Villigen in the frame of the European Fusion Technology Programme and in cooperation with industry. This current lead consists of an HTS part that covered the temperature range from 4.5 to 70 K and a conventional part, making the connection to room temperature. Because the HTS part had no ohmic losses and poor thermal conduction, the refrigerator power necessary for cooling the current lead was reduced drastically. The saving factor could be calculated to be 5.4 at zero current and 3.7 at 68 kA. The current lead could even be operated at 80 kA and with respect to safety criteria of ITER, a complete loss of He flow was simulated showing that the HTS current lead could hold a current of 68 kA for 6 min without active cooling. These results demonstrate that today existing HTS materials can be used in ITER for current leads or bus bar systems.For fusion machines beyond ITER, the development of an HTS fusion conductor would be the key to operate the complete magnet system at higher temperatures. The option of developing fusion conductors based on Bi-2223 and YBCO are briefly discussed. For a success of such conductors, the AC loss optimisation is crucial.     

基于RELAP5的双功能液态锂铅实验包层模块安全分析

利用嵌入了液态锂铅（LiPb）的热工水力子模块的系统程序RELAP5/MOD3,对双功能液态锂铅（DFLL）实验包层模块（TBM）的安全特性进行评价。对DFLL-TBM及其辅助冷却系统的稳态运行工况、预期运行事件和相关事故工况进行了建模、计算和分析。计算结果表明,稳态运行时第一壁（FW）结构材料表面最高温度低于允许值550 ℃。事故工况下氦气泄漏引起的ITER真空室（VV）、窗口设备室（port cell）以及托卡马克冷却水系统大厅拱顶（TCWS vault）的增压均低于ITER要求的限值0.2 MPa。实验包层钢结构不会熔化且可通过辐射换热有效地导出衰变余热。DFLL-TBM的设计可满足ITER对其热工水力安全方面的要求。     

LOCA analysis for Korean helium cooled solid breeder TBM

One of the major ITER goals is test blanket module (TBM) program which is for the demonstration of the breeding capability that would lead to tritium self-sufficiency in a reactor and the extraction of high-grade heat suitable for electricity generation under the ITER fusion environment. While the engineering design of Korean helium cooled solid breeder (HCSB) TBM and its ancillary systems has been performed, a safety assessment on different possible accident scenarios should be carried out for the purpose of licensing. In this paper, accident analyses for several loss of coolant accident (LOCA) cases were performed in order to assess safety aspects of the TBM design using RELAP5/MOD3.2. Since the TBM forms a loop with helium cooling system (HCS) which is one of ancillary systems required for removing heat deposited in the TBM by neutron wall loading and surface heat flux from plasma, it is necessary to model the complete loop for accident analysis. In this study, the helium passage including the TBM and HCS was nodalized for each accident scenario. The TBM and HCS components were modeled as the associated heat structures provided by RELAP5 to include heat transfer across solid boundaries. Based on computational results it was found that current design of the TBM is robust from the safety point of view.     

Thermo-hydraulic and structural analysis for finger-based concept of ITER blanket first wall

The blanket first wall is one of the main plasma facing components in ITER tokamak. The finger-typed first wall was proposed through the current design progress by ITER organization. In this concept, each first wall module is composed of a beam and twenty fingers. The main function of the first wall is to remove efficiently the high heat flux loading from the fusion plasma during its operation. Therefore, the thermal and structural performance should be investigated for the proposed finger-based design concept of first wall. The various case studies were performed for a unit finger model considering different loading conditions. The finite element model was made for a half of a module using symmetric boundary conditions to reduce the computational effort. The thermo-hydraulic analysis was performed to obtain the pressure drop and temperature profiles. Then the structural analysis was carried out using the maximum temperature distribution obtained in thermo-hydraulic analysis. Finally, the transient thermo-hydraulic analysis was performed for the generic first wall module to obtain the temperature evolution history considering cyclic heat flux loading with nuclear heating. After that, the thermo-mechanical analysis was performed at the time step when the maximum temperature gradient was occurred. Also, the stress analysis was performed for the component with a finger and a beam to check the residual stress of the component after thermal shrinkage assembly.     

Preliminary thermal-hydraulic and safety analysis of China DFLL-TBM system

China has proposed the dual-functional lithium-lead (DFLL) tritium breeding blanket concept for testing in ITER as a test blanket module (TBM), to demonstrate the technologies of tritium self-sufficiency, high-grade heat extraction and efficient electricity production which are needed for DEMO and fusion power plant. Safety assessment of the TBM and its auxiliary system should be conducted to deal with ITER safety issues directly caused by the TBM system failure during the design process. In this work, three potential initial events (PIEs) – in-vessel loss of helium (He) coolant and ex-vessel loss of He coolant and loss of flow without scram (LOFWS) – were analyzed for the TBM system with a modified version of the RELAP5/MOD3 code containing liquid lithium-lead eutectic (LiPb). The code also comprised an empirical expression for MHD pressure drop relevant to three-dimensional (3D) effect, the Lubarsky–Kaufman convective heat transfer correlation for LiPb flow and the Gnielinski convective heat transfer correlation for He flow. Since both LiPb and He serve as TBM coolants, the LiPb and He ancillary cooling systems were modeled to investigate the thermal-hydraulic characteristic of the TBM system and its influence on ITER safety under those accident conditions. The TBM components and the coolants flow within the TBM were simulated with one-dimensional heat structures and their associated hydrodynamic components. ITER enclosures including vacuum vessel (VV), port cell and TCWS vault were also covered in the model for accident analyses. Through this best estimate approach, the calculation indicated that the current design of DFLL-TBM and its auxiliary system meets the thermal-hydraulic and safety requirements from ITER.     

Overview of results from Wendelstein7-X HTS current lead testing

The High Temperature Superconductor (HTS) current leads (CL) for the Wendelstein 7-X stellarator (W7-X) with a maximum current of 18.2 kA are designed and manufactured by the Karlsruhe Institute of Technology (KIT). In addition the acceptance tests of the W7-X HTS CLs are performed at KIT. Therefore the existing TOSKA facility has been extended by a test cryostat connected to the main vacuum vessel. After the extensive prototype CL test campaigns in 2010 the final acceptance tests of 14 series CLs started in 2011. The estimated completion of the routine test campaign is in December 2012. The main parts of each acceptance test are the determination of the heat load at the 4.5 K level, of the necessary 50 K He mass flow rate through the heat exchanger as well as the simulation of a loss of flow accident of the 50 K He mass flow at full current (18.2 kA). The tests also include a long-time operation at the maximum current of 18.2 kA to demonstrate the steady state operation capability of the HTS CLs. In the present paper an overview of all conducted HTS CL acceptance tests is given. The results for the different CLs are summarized and compared to the specifications.     

Study of electromagnetic interference on quench detecting system of HTS current leads for EAST

High temperature superconducting (HTS) material B-2223/Ag-Au has been used for EAST poloidal field (PF) coil current leads for reducing construction and operation cost of cryogenic system. The quench propagation velocity of HTS superconducting material is several orders of magnitude lower than that of normal low temperature current leads. It is difficult to detect weak signal of quench which is easily influenced by strong electromagnetic interference (EMI). In this paper, the sources of EMI from quench detecting system of high temperature current leads have been introduced. And we have chosen reasonable methods for good transformation and protection on the basis of electromagnetic compatibility simulation diagnosis experiments. Recent experimental results showed that the restraint of EMI has been achieved and has met the requirements of experiment.     

Thermal Analysis of Power Thyristor Used in ITER Poloidal Field Converter

As a key component of ITER PF converter module, thyristor conducting large current produces a lot of power loss, therefore it is meaningful to study its heat transfer characteristic for improving the performance of PF converter. This paper presents the thermal analysis of the thyristor. A 3D finite-element model with multi material layers is built and simulated in steady state operation. A special temperature rise test scheme is designed and done to verify the analysis result. The simulation is well in compliance with the test result. The modeling method presented in this paper is proved to be practical in thyristor thermal analysis. The proposed measurement method in temperature rise test is also of valuable reference for thermal testing of power semiconductor devices.