Thermo-hydraulic analysis for CN HCCB TBM regarding ITER new baseline scenario

IF 2 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Fusion Engineering and Design Pub Date : 2025-09-11 DOI:10.1016/j.fusengdes.2025.115427

Xinghua Wu, Qixiang Cao, Fengchao Zhao, Long Zhang, Xiaoyu Wang

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引用次数: 0

Abstract

Demonstration of tritium breeding blanket technique is one of the engineering goal of International Thermonuclear Experimental Reactor (ITER). Among the different blanket concepts, China decided to develop Helium-cooled Ceramic-Breeder Test Blanket Module (HCCB TBM) and tested in ITER specific TBM port. Currently CN HCCB TBM design is under preliminary design (PD) phase. In 2023, regarding some engineering and technical issues, ITER proposed to modify previous baseline scenario, with less than 1% of total neutron fluence in DT-1 phase, and reduced plasma pulse parameters. Regarding the testing goals of TBM programme, additional electrical heater was proposed to put inside TBM module, so as to achieve relatively high temperature for tritium release. In this paper, steady-state and transient thermo-hydraulic analysis was performed for HCCB TBM typical component, and the overall temperature distribution was achieved, which will be input for structural analysis and dynamic tritium transport analysis in the next step.

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CN HCCB TBM在ITER新基线情景下的热水力分析

氚增殖包层技术的论证是国际热核实验反应堆（ITER）的工程目标之一。在不同的包层概念中，中国决定发展氦冷却陶瓷-增殖器测试包层模块（HCCB TBM）并在ITER特定的TBM端口进行测试。目前CN HCCB TBM设计处于初步设计（PD）阶段。2023年，由于一些工程技术问题，ITER提出修改之前的基线情景，将DT-1相中子总通量控制在1%以下，并降低等离子体脉冲参数。针对TBM方案的测试目标，提出在TBM模块内增加电加热器，以达到较高的氚释放温度。本文对HCCB TBM典型部件进行了稳态和瞬态热水力分析，获得了整体温度分布，为下一步的结构分析和动态氚输运分析提供数据。

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.