ITER will use beryllium as a plasma-facing material in the main chamber, covering a total surface area of about 620 m${}^2$. Given the importance of beryllium erosion and co-deposition for tritium retention in ITER, significant efforts have been made to understand the behaviour of beryllium under fusion-relevant conditions with high particle and heat loads. This paper provides a comprehensive report on the state of knowledge of beryllium behaviour under fusion-relevant conditions: the erosion mechanisms and their consequences, beryllium migration in JET, fuel retention and dust generation. The paper reviews basic laboratory studies, advanced computer simulations and experience from laboratory plasma experiments in linear simulators of plasma–wall interactions and in controlled fusion devices using beryllium plasma-facing components. A critical assessment of analytical methods and simulation codes used in beryllium studies is given. The overall objective is to review the existing set of data with a broad literature survey and to identify gaps and research needs to broaden the database for ITER.

ITER将在主室中使用铍作为面向等离子体的材料，覆盖约620 m的总表面积${}^{\mathrm{2个}}$。鉴于铍腐蚀和共沉积对于ITER中retention保留的重要性，已经做出了巨大的努力来了解铍在与熔融相关的条件下具有高颗粒和高热负荷的行为。本文提供了有关与聚变有关的条件下铍行为的知识状态的综合报告：腐蚀机理及其后果，铍在JET中的迁移，燃料滞留和粉尘产生。本文回顾了基础的实验室研究，先进的计算机模拟以及在等离子-壁相互作用的线性仿真器中以及在使用铍等离子处理元件的受控聚变设备中进行的实验室等离子实验的经验。对铍研究中使用的分析方法和模拟代码进行了严格的评估。