Tritium release scenarios in the IRPR system of DEMO: Safety assessment and mitigation strategies

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

Fusion Engineering and Design Pub Date : 2025-10-11 DOI:10.1016/j.fusengdes.2025.115477

Samuele Meschini, Raffaella Testoni

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引用次数: 0

Abstract

This study investigates potential tritium release scenarios in the Isotope Rebalancing and Protium Removal (IRPR) system of DEMO fuel cycle, focusing on failures in the Temperature Swing Absorption (TSA) unit, including its feed tank. Two accident cases are considered, based on the postulated initiating events identified by the FFMEA: (1) a guillotine break in the pipeline supplying the IRPR system and (2) a pipe rupture in the TSA unit. MELCOR 1.8.6 for fusion is used to simulate the dynamics of the gases (hydrogen, air, inert gases), the resulting pressurization of the enclosure volumes, and the accumulation of hydrogen isotopes. A parametric study evaluates the impact of design choices like feed tank configuration and glovebox volume, informing the design of the system to enhance safety. The analysis includes cases where safety systems such as detritiation or isolation fail to operate correctly. The findings support ongoing safety assessments for DEMO and inform strategies to improve safety in fusion power plants.

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DEMO的IRPR系统中的氚释放情景：安全评估和缓解策略

本研究探讨了DEMO燃料循环中同位素再平衡和除Protium （IRPR）系统中可能出现的氚释放情况，重点研究了变温吸收（TSA）装置（包括其进料罐）的故障。基于FFMEA确定的假设启动事件，考虑了两种事故情况：(1)供应IRPR系统的管道中的断头台断裂；(2)TSA单元的管道破裂。用于聚变的MELCOR 1.8.6用于模拟气体（氢气，空气，惰性气体）的动力学，由此产生的封闭体积的加压以及氢同位素的积累。参数化研究评估了设计选择的影响，如进料箱配置和手套箱体积，为系统设计提供信息，以提高安全性。分析包括安全系统，如防蚀或隔离不能正确操作的情况。研究结果支持正在进行的DEMO安全评估，并为提高核聚变发电厂的安全性提供战略信息。

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.