The Heating & Current Drive System of Divertor Tokamak Test (DTT)

G. Granucci, G. Ravera, A. Bruschi, S. Ceccuzzi, P. Agostinetti, S. Garavaglia, A. Romano, A. Ferro
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引用次数: 3

Abstract

Appropriate disposal of the non-neutronic energy and particle exhaust in a reactor is universally recognized as one of the high priority challenges for the exploitation of fusion as an energy source. The Divertor Tokamak Test (DTT) facility will be built to study a solution suitable for the power exhaust in conditions relevant for DEMO. The tokamak will reach the needed condition of 15 MW/m power flow to the divertor by coupling up to 45 MW of additional power to the plasma. The selected Heating Systems to achieve this goal are Electron Cyclotron Heating (ECH), Ion Cyclotron Heating (ICH) and Negative Neutron Beam Injector (NNBI). The power systems will be installed in two stages: at a first stage 16 MW of ECRH power, 3 MW of ICRH and one 7.5 MW NNBI injector will be installed, making the DTT plasma suitable for relevant experiment at 6T, 4MA configuration. The EC system relies on a 170 GHz, 1 MW gyrotron, similar to those developed for ITER, while for the power transmission a Quasi Optical approach has been chosen, where a multi-beam mirrors will be installed under vacuum to reduce the overall transmission losses below 10%. The power will be injected exploiting independent front-steering antennas capable to steer in real-time all the beams. The module of the ICRH system will be based on transmitters, capable of a wide frequency range (60-90 MHz), connected to two movable antennas inserted in the equatorial ports of DTT. The choice of the antenna type will be based on reliability (i.e. power density) rather than on its performance in terms of peak coupled power. Fast variations of the antenna loading, as the one expected in presence of ELM, will be compensated exploiting an external matching system. The NNBI will be based on two RF plasma sources capable to produce a negative ion current that will be accelerated by a grids system up to 400 keV. The designed injector will reflect the experience gained in SPIDER and MITICA, with modifications aiming to a simplified and well performing system. The paper describes the main characteristics of the design of DTT additional heating system, that will be one of the most powerful between the tokamaks of the next generation.
导向器托卡马克试验(DTT)加热及电流驱动系统
在反应堆中合理处理非中子能量和粒子废气被普遍认为是利用核聚变作为一种能源的高优先级挑战之一。将建立转向器托卡马克测试(DTT)设施,以研究适用于与DEMO相关条件下的动力排气的解决方案。托卡马克将通过向等离子体耦合高达45兆瓦的额外功率来达到向转向器输送15兆瓦/米功率的所需条件。实现这一目标的加热系统有电子回旋加热(ECH)、离子回旋加热(ICH)和负中子束注入器(NNBI)。电力系统将分两个阶段安装:第一阶段将安装16兆瓦的ECRH功率,3兆瓦的ICRH功率和一个7.5兆瓦的NNBI注入器,使DTT等离子体适合在6T, 4MA配置下进行相关实验。EC系统依赖于一个170 GHz, 1 MW的回旋管,类似于为ITER开发的回旋管,而电力传输则选择了准光学方法,其中将在真空下安装多光束反射镜,以将总体传输损耗降低到10%以下。能量的注入将利用能够实时引导所有波束的独立前转向天线。ICRH系统的模块将基于发射机,具有宽频率范围(60- 90mhz),连接到插入DTT赤道端口的两个可移动天线。天线类型的选择将基于可靠性(即功率密度),而不是其峰值耦合功率方面的性能。天线载荷的快速变化,如预期的ELM存在时,将利用外部匹配系统进行补偿。NNBI将基于两个能够产生负离子电流的射频等离子体源,该负离子电流将由网格系统加速至400 keV。设计的注入器将反映在SPIDER和MITICA中获得的经验,并进行修改,旨在简化和性能良好的系统。本文介绍了下一代托卡马克中最强大的DTT附加加热系统设计的主要特点。
本文章由计算机程序翻译,如有差异,请以英文原文为准。
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