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。     

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。     

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。     

ITER中的电子回旋波电流驱动模拟

通过将相对论Fokker-Planck方程与波迹方程联合求解,对ITER(国际热核实验反应堆)参数下的电子回旋波电流驱动进行了数值模拟。结果表明,当波的环向发射角度不太大时,波功率沉积将发生在ITER的强场侧。当环向发射角度为21°时,电子回旋波的能量在等离子体中心区域被吸收并驱动起等离子体中心区域的电流。当发射角度变大时,电子回旋波将在弱场侧沉积功率。当发射角度为20°～30°时,能够驱动归一化的径向位置(r/a)小于0.35区域内的等离子体电流,并有较高的电流驱动效率。     

可控核聚变与国际热核实验堆（ITER）计划

介绍了我国能源的基本隋况,核聚变能和可控核聚变的基本原理,以及国际热核聚变实验堆ITER的历史与现状。对我国磁约束核聚变的研究发展历程做了简要的回顾。     

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计划取得两项重要进展

正【本刊2014年9月综合报道】近期,国际热核聚变实验堆(ITER)计划已取得两项重要进展,即完成托卡马克厂房混凝土基础底板的浇筑以及将首批系统部件运抵场区。建成托卡马克厂房基础底板当地时间2014年8月27日18时许,工作人员历时12小时完成了ITER托卡马克厂房混凝土基础底板(B2混凝土基础底板)     

R＆D on Resistive Heat Exchangers for HTS High Rated Current Leads

The HTS current leads of superconducting magnets for large scale fusion devices and high energy particle colliders can reduce the power consumption for cooling ...     

ITER屏蔽块冷却通道热工水力分析与设计改进

屏蔽块是国际热核聚变实验堆(ITER)的重要组成部分之一,其结构设计尤其是内部冷却剂通道设计将直接影响屏蔽块的冷却效果及模型的安全性。运用计算流体力学(CFD)软件对屏蔽块模型进行热工水力计算,根据计算结果提出改进建议,并对改进后的模型重新进行数值分析。结果表明:屏蔽块的进出口段压降减小,冷却剂流动总压降减小;主要冷却支管的流量分配和流速分布更加均匀;屏蔽块表面局部最高温度大大降低,可避免由于冷却不均造成模块热应力过大。     

Design modification and optimization of the ITER cooling water system

ITER (Latin for “the way”), the largest fusion experimental reactor in the world, is designed to demonstrate the technological feasibility of nuclear fusion energy conversion, at plant scale, from high temperature deuterium-tritium plasma using the Tokamak magnetic confinement arrangement.ITER will have a large vacuum vessel that hosts the plasma facing components. These components include the blanket and the divertor that will operate at temperatures, heat loads, and neutron flux higher than those reached in a nuclear fission power plant reactor.One of the main critical issues of the ITER reactor is the design of the cooling water system to transfer the heat generated in the plasma to the in-vessel components and the heat loads from the auxiliary systems to the environment.This paper describes the current ITER cooling water system and recent design modifications and optimizations.     

Lessons Drawn from ITER and Other Fusion International Collaborations

The international character of fusion research and development is described, with special emphasis on the ITER (International Thermonuclear Experimental Reactor) joint venture. The history of the ITER collaboration is traced. Lessons drawn that may prove useful for future ventures are presented.     

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。     

Stability estimations and quench simulations for the ITER conductors

Quench simulations and stability estimations for the International Thermonuclear Experimental Reactor (ITER) are discussed. Especially numerical issues and associated benchmark actions are summarized. Satisfactory agreement between the various codes from the 4 ITER parties is now obtained after numerical convergence problems have been resolved. However, these require confirmation by experiments on relevant conductor geometries. In multistage cables, a non-uniform current distribution within the cable affects the stability of the conductor. A possible mechanism for the non-uniform distribution is flux loops between the strands or cable substages as the current is ramped up or down. A preliminary estimation of stability with non-uniform current distribution is also discussed.     

Safety and Environmental Activities for ITER

This paper will summarize highlights of the safety approach and discuss the ITER EDA safety activities. The ITER safety approach is driven by three major objectives: (1) Enhancement or improvement of fusion's intrinsic safety characteristics to the maximum extent feasible, which includes a minimization of the dependence on dedicated safety systems; (2) Selection of conservative design parameters and development of a robust design to accommodate uncertainties in plasma physics as well as the lack of operational experience and data; and (3) Integration of engineered mitigation systems to enhance the safety assurance against potentially hazardous inventories in the device by deploying well-established nuclear safety approaches and methodologies tailored as appropriate for ITER.     

Scaling of the Density Peak with Pellet Injection in ITER

Scalings of the density peak and pellet penetration length in ITER are developed based on simulations using 1.5D BALDUR integrated predictive modeling code. In these simula- tions, the pellet ablation is described by the Neutral Gas Shielding (NGS) model with grad-B drift effect taken into account. The NGS pellet model is coupled with a plasma core transport model, which is a combination of an MMM95 anomalous transport model and an NCLASS neoclassical transport model. The BALDUR code with a combination of MMM95 and NCLASS models, together with the NGS model, is used to simulate the time evolution of plasma current, ion and electron temperatures, and density profiles for ITER standard type I ELMy H-mode discharges during the pellet injection. As a result, the scaling of the density peak and pellet penetration length at peak density can be established using this set of predictive simulations that covers a wide range of ITER plasma conditions and pellet parameters. The multiple regression technique is utilized in the development of the scalings. It is found that the scaling for density at center is sensitive to both the plasma and pellet parameters; whereas the scalings for density and location of the additional peak are sensitive to the pellet parameters only.     

The ITER Safety Approach for External Events

ITER will be the first large-scale tokamak to be designed as a nuclear facility to provide public protection from external hazards such as earthquakes. The design approach for such events has been developed consistent with ITER's moderate hazards and overall safety approach on a basis of the ITER site assumptions. Seismic design is described including selection of ground motions for design purposes, seismic safety requirements, and the seismic classification scheme. The results of preliminary seismic assessments are summarized including the potential for seismically induced plasma vertical displacement events (VDE). Finally, potential facility modifications available to deal with site-specific external hazards are suggested. At the Detailed Design Report stage of the Engineering Design Activity (EDA), it is concluded that ITER has been designed to deal with the site design assumptions for earthquakes and can be designed to safety cope with a range of site-specific external hazards with modest changes to the facility.