Installation, thermal curing, qualification testing of divertor and position control coils in ADITYA-U tokamak

IF 1.9 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY Fusion Engineering and Design Pub Date : 2024-11-22 DOI:10.1016/j.fusengdes.2024.114734

Rohit Kumar , Vaibhav Ranjan , Harshita Raj , Sharvil Patel , K. Sathyanarayana , Kaushal Patel , Kumarpal Jadeja , R.L. Tanna , J. Ghosh

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Abstract

The former ADITYA, a medium-sized tokamak with a limiter configuration was upgraded to ADITYA-U tokamak with divertor configuration. Two pairs of new divertor coils, a single pair of auxiliary divertor coils and position control coils have been introduced in ADITYA-U tokamak to achieve shaped plasma operation using the existing Toroidal field, Ohmic transformer and vertical field coils. Currently, copper-based continuous transposed conductor (CTC) has been introduced for in-situ winding of the coils. Coil insulation materials are selected to withstand high current (15 kA), high voltage (5 kV) and sustain high temperature (120 °C) during the experiment. The primary challenge was to install the coil with the bus bar using the same conductor without any joints. The design of new divertor coils mainly includes electrical and thermal considerations within the limited space available for installation. A detailed description of the installation of coils, insulation fabrication, insulation curing process and testing of the divertor coils is presented in this paper.

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ADITYA-U 托卡马克中分流器和位置控制线圈的安装、热固化和鉴定测试

ADITYA-U托卡马克的前身是采用限幅器结构的中型托卡马克，现已升级为采用岔流器结构的ADITYA-U托卡马克。ADITYA-U 托卡马克引入了两对新的分流线圈、一对辅助分流线圈和位置控制线圈，以利用现有的环形场、欧姆变压器和垂直场线圈实现定形等离子体运行。目前，已采用铜基连续换位导体（CTC）对线圈进行就地绕组。线圈绝缘材料的选择是为了在实验过程中承受大电流（15 kA）、高电压（5 kV）和持续高温（120 °C）。首要挑战是使用相同导体将线圈与母线安装在一起，且不留任何接头。在有限的安装空间内，新型分流线圈的设计主要包括电气和热方面的考虑。本文详细介绍了线圈的安装、绝缘层的制作、绝缘层的固化过程以及分流线圈的测试。

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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来源期刊

Fusion Engineering and Design 工程技术-核科学技术

CiteScore

3.50

自引率

23.50%

发文量

275

审稿时长

3.8 months

期刊介绍： The journal accepts papers about experiments (both plasma and technology), theory, models, methods, and designs in areas relating to technology, engineering, and applied science aspects of magnetic and inertial fusion energy. Specific areas of interest include: MFE and IFE design studies for experiments and reactors; fusion nuclear technologies and materials, including blankets and shields; analysis of reactor plasmas; plasma heating, fuelling, and vacuum systems; drivers, targets, and special technologies for IFE, controls and diagnostics; fuel cycle analysis and tritium reprocessing and handling; operations and remote maintenance of reactors; safety, decommissioning, and waste management; economic and environmental analysis of components and systems.