与液体Pb-16Li接触的流道镶块中氧化锆的电导率

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

Fusion Engineering and Design Pub Date : 2025-06-12 DOI:10.1016/j.fusengdes.2025.115260

Jaroslav Kekrt, Roman Petráš, Adéla Gottfriedová, Vojtěch Svoboda, Ladislav Vála

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引用次数: 0

摘要

流动通道嵌套（FCI）作为功能部件正在开发用于双冷却剂锂铅繁殖毯。由于其目的是电去耦液体铅锂的主要流动和覆盖层的钢结构，因此了解FCI材料的电学特性至关重要。为此，在氧化陶瓷- ZrO2上进行了一系列的直接电流测量，以氧化钇稳定氧化锆的形式。最初，在惰性无氧气氛中，在240-520°C的温度范围内测定电导率，使用平面样品两侧的两个工作电极。随后，在其他相同的实验条件下，使用液态Pb-16Li代替其中一个电极来测定样品的电学性质。在给定温度范围内，材料的导电性远低于流动通道插入件的要求极限。

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Electrical conductivity of zirconia for flow channel inserts in contact with liquid Pb-16Li

Flow channel inserts (FCI) as functional components are being developed for the Dual Coolant Lithium Lead breeding blanket. Since the aim is to electrically decouple the main flow of liquid lead lithium and a steel structure of the blanket, knowledge of electrical properties of the FCI materials is crucial. For this purpose, series of direct electrical current measurements were performed on oxide ceramic – ZrO₂ in a form of yttria stabilized zirconia. Initially, the electrical conductivity was determined in temperature range of 240-520 °C in an inert oxygen-free atmosphere using two working electrodes on the opposite sides of a flat sample. Subsequently, the electrical properties of the samples were determined using liquid Pb-16Li as a substitution of one of the electrodes, under otherwise identical experimental conditions.

The electrical conductivity of materials in the given temperature range is well below the required limit of the Flow channel inserts.

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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.