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A component test facility based on the spherical tokamak

机译：基于球形托卡马克的组件测试设备

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Recent experiments (Synakowski et al 2004 Nucl. Fusion 43 1648, Lloyd et al 2004 Plasma Phys. Control. Fusion 46 13477) on the Spherical Tokamak (or Spherical Torus, ST) (Peng 2000 Phys. Plasmas 7 1681) have discovered robust plasma conditions, easing shaping, stability limits, energy confinement, self-driven current and sustainment. This progress has encouraged an update of the plasma conditions and engineering of a Component Test Facility (CTF), (Cheng 1998 Fusion Eng. Des. 38 219) which is a very valuable step in the development of practical fusion energy. The testing conditions in a CTF are characterized by high fusion neutron fluxes Gamma(n) approximate to 8.8 x 10(13) n s(-1) cm(-2) ('wall loading' W-L approximate to 2 MW m(-2)), over size-scale > 10(5) cm(2) and depth-scale > 50 cm, delivering > 3 accumulated displacement per atom per year ('neutron fluence' > 0.3 MW yr(-1) m(-2)) (Abdou et al 1999 Fusion Technol. 29 1). Such conditions are estimated to be achievable in a CTF with R-0 = 1.2 m, A = 1.5, elongation similar to 3, I-p similar to 12 MA, B-T similar to 2.5 T, producing a driven fusion burn using 47 MW of combined neutral beam and RF heating power. A design concept that allows straight-line access via remote handling to all activated fusion core components is developed and presented. The ST CTF will test the lifetime of single-turn, copper alloy centre leg for the toroidal field coil without an induction solenoid and neutron shielding and require physics data on solenoid-free plasma current initiation, ramp-up to and sustainment at multiple megaampere level. A systems code that combines the key required plasma and engineering science conditions of CTF has been prepared and utilized as part of this study. The results show high potential for a family of relatively low cost CTF devices to suit a range of fusion engineering and technology test missions.
机译：最近在球形托卡马克（或Spherical Torus，ST）（Peng 2000 Phys.Plasmas 7 1681）上进行的实验（Synakowski等2004 Nucl.Fusion 43 1648，Lloyd等2004 Plasma Phys.Control.Fusion 46 13477）发现了坚固的等离子体条件，缓和成形，稳定性极限，能量限制，自驱动电流和维持性。这一进展鼓励了等离子体条件的更新和组件测试设施（CTF）的改造（Cheng 1998 Fusion Eng。Des。38 219），这是开发实用聚变能中非常有价值的一步。 CTF中的测试条件以高聚变中子通量Gamma（n）约8.8 x 10（13）ns（-1）cm（-2）为特征（“壁负载” WL约2 MW m（-2） ），尺寸尺度> 10（5）cm（2），深度尺度> 50 cm，每年每原子传递> 3的累积位移（``中子注量''> 0.3 MW yr（-1）m（-2） （Abdou et al 1999 Fusion Technol.29 1）。据估计，在C-0中，R-0 = 1.2 m，A = 1.5，伸长率近似于3，Ip近似于12 MA，BT近似于2.5 T，可以实现这种条件，从而使用47 MW的组合中性线产生了驱动的聚变燃烧光束和射频加热功率。提出并提出了一种设计概念，该概念允许通过远程处理直接访问所有激活的融合核心组件。 ST CTF将测试无感应螺线管和中子屏蔽的环形磁场线圈的单匝铜合金中心脚的寿命，并要求有关无螺线管的等离子电流启动，上升并维持在数兆安级别的物理数据。结合关键要求的等离子体和CTF工程科学条件的系统代码已准备就绪，并已用作本研究的一部分。结果表明，一系列成本相对较低的CTF设备具有很高的潜力，可满足一系列融合工程和技术测试任务的需要。

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