Thermal hydraulic analysis of in-vessel loss of coolant accident for the EAST lower divertor primary heat transfer system

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

Fusion Engineering and Design Pub Date : 2024-08-03 DOI:10.1016/j.fusengdes.2024.114613

Jinxuan Zhou , Jiansheng Hu , Bin Guo , Lei Yang , Weibao Li

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Abstract

Uneven heat load distribution on the divertor during the high power long-pluse discharge of the Experimental Advanced Superconducting Tokamak (EAST) leads to hot spot phenomena, potentially causing the Plasma Facing Component (PFC) material melting, flaking, and even penetration, which may trigger the in-vessel loss of coolant accident (LOCA). Continuous coolant intrusion could damage vacuum equipment, while flash evaporation may increase vacuum pressure, posing a potential threat to the safety of the device operation. In this research, the RELAP5/MOD3.4 program was employed to develope a model of the lower divertor primary heat transfer system (PHTS). Steady state analysis was conducted to obtain the key parameters of the system in comparison with the design parameters, and the results showed good consistency. Thermal-hydraulic analysis of the in-vessel LOCA is performed based on the design condition, quantitatively investigating the evolution of the breach discharge flow rate and vacuum pressure. An additional pneumatic isolation valve and check valve are proposed as an accident mitigation scheme, and the effectiveness is evaluated to provide a reference for the upgrade of EAST lower divertor PHTS.

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EAST 下部分流器一次传热系统舱内冷却剂损失事故的热工水力分析

先进超导实验托卡马克（EAST）在高功率长普放电过程中，分流器上的热负荷分布不均会导致热点现象，有可能造成等离子体面组件（PFC）材料熔化、剥落甚至穿透，从而引发腔内冷却剂损失事故（LOCA）。持续的冷却剂侵入可能会损坏真空设备，而闪蒸则可能会增加真空压力，对装置的运行安全构成潜在威胁。在这项研究中，采用 RELAP5/MOD3.4 程序开发了下转发器初级传热系统（PHTS）模型。通过与设计参数的对比，进行了稳态分析以获得系统的关键参数，结果显示出良好的一致性。根据设计条件对舱内 LOCA 进行了热液压分析，定量研究了裂口排放流量和真空压力的变化。提出了附加气动隔离阀和止回阀的事故缓解方案，并对其有效性进行了评估，为 EAST 下分流器 PHTS 的升级提供参考。

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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.