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Дата издания
01.12.2023На каком языке издана
ISSN
0920-3796Область исследования
Тип исследования
Аннотация (краткое описание)
The design of future fusion reactors involves the production of tritium inside the breeder blanket. The most promising material for solid breeder blankets is a two-phase lithium ceramic containing orthosilicate Li4SiO4 (LOS) and metatitanate Li2TiO3 (LMT) of lithium in various proportions. Tritium is formed in lithium under neutron irradiation by the reaction 6Li(n,α)T. Further, this tritium is extracted from the blanket with a purge gas and returned to the fusion zone, realizing the concept of a closed fusion cycle. Irradiation under fission reactor conditions is still one of the few available methods for estimating the parameters of tritium generation and release from lithium-containing materials in the "in-situ" mode. This paper presents the results of experiments on neutron irradiation of two-phase lithium ceramics of various ratios (LOS + 35 mol. % LMT (pebble size 250–1250 μm), LOS + 35 mol. % LMT (pebble size 500–710 μm) and LOS + 25 mol. % LMT (pebble size 500–710 μm) at the WWR-K research reactor. Irradiation of each batch of samples lasted from 5 to 22 days. The experiments were carried out by the vacuum extraction method. This paper describes the main methodological aspects of the studies, namely the technical features of four irradiation campaigns, the sequence and scope of the studies. A comparison is also made of the initial sections of reactor experiments for all campaigns, where the reactor was sequentially brought to power, according to which the parameters of the Arrhenius dependence of the effective tritium diffusion coefficients were estimated. © 2023 Elsevier B.V.DOI
10.1016/j.fusengdes.2023.114035Тип публикаций
Практическое значение
The study provides important experimental data on tritium generation and release in lithium-based ceramic breeder materials intended for future fusion reactors. The results help optimize the composition and microstructure of breeder blankets, improving tritium production efficiency and supporting the development of a closed fuel cycle in fusion energy systems. In addition, the estimated tritium diffusion parameters contribute to the design, safety assessment, and performance prediction of fusion reactor blanket materials under neutron irradiation conditions.Квартиль
Источник квартиля
SCImago (Scopus)Impact Factor
3.5Количество цитирований
18Данный документ располагается в коллекциях
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