ITER容器内线圈陶瓷绝缘断口的设计与鉴定

IF 1.9 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Fusion Engineering and Design Pub Date : 2025-04-22 DOI:10.1016/j.fusengdes.2025.115112

Davide Macioce , Alexander Vostner , Cristina Jimenez Aguinaga , Xabier Gomez Mitxelena , Miguel Angel Lagos Gomez , Igor Rueda Salazar , Roberto San Martín Ruiz , Ignacio Aviles Santillana

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引用次数: 0

摘要

热核实验堆容器内线圈（IVC）是水冷线圈，由矿物绝缘铜导体组成，并包裹在不锈钢护套内。该系统安装在热核实验堆真空容器内，由高压馈入装置为铜导体提供电力和冷却水。每个线圈都单独冷却，并在舱外区域装有专用的不锈钢进口/出口管道。IVC 绝缘断路器 (IB) 在辐射环境和高达 240 °C 的温度下运行。IVC 绝缘断路器 (IB) 在辐射环境和高达 240 °C 的温度下运行。它应提供高达 2.4 kV 的电气绝缘，并应承受高达 44 bar 的冷却水压力以及与热核实验堆 IVC 运行相关的机械和热负荷。

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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Design and qualification of a ceramic insulating break for ITER In-Vessel Coils

The ITER In-Vessel Coils (IVC) are water-cooled coils, composed by a mineral-insulated copper conductor, enclosed in a stainless-steel jacket. The system is installed inside the ITER Vacuum Vessel, and it is fed by high-voltage feedthroughs that provide electrical power and cooling water to the copper conductors. Each coil is cooled individually and has dedicated inlet/outlet stainless-steel pipes in the ex-vessel area. To electrically separate the IVC from the ITER cooling water system, an electrical insulating break is needed.

The IVC Insulating Break (IB) is operated in a radiation environment and at temperatures up to 240 °C. It shall provide electrical insulation up to 2.4 kV and shall withstand cooling water pressure up to 44 bar as well as mechanical and thermal loads related to the ITER IVC operations.

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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.