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首页> 外文期刊>Nuclear Engineering and Design >An integrated model of tritium transport and corrosion in Fluoride Salt-Cooled High-Temperature Reactors (FHRs) - Part I: Theory and benchmarking
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An integrated model of tritium transport and corrosion in Fluoride Salt-Cooled High-Temperature Reactors (FHRs) - Part I: Theory and benchmarking

机译：氟化盐冷却高温反应堆（FHRs）中transport运输和腐蚀的集成模型-第一部分：理论和基准

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The Fluoride Salt-Cooled High-Temperature Reactor (FHR) is a pebble bed nuclear reactor concept cooled by a liquid fluoride salt known as "flibe" ((LiF)-Li-7-BeF2). A model of TRITium Diffusion EvolutioN and Transport (TRIDENT) was developed for use with FHRs and benchmarked with experimental data, TRIDENT is the first model to integrate the effects of tritium production in the salt via neutron transmutation, with the effects of the chemical redox potential, tritium mass transfer, tritium diffusion through pipe walls, tritium uptake by graphite, selective chromium attack by tritium fluoride, and corrosion product mass transfer. While data from a forced-convection polythermal loop of molten salt containing tritium did not exist for comparison, TRIDENT calculations were compared to data from static salt diffusion tests in flibe and flinak (0.465LiF-0.115NaF-0.42KF) salts. In each case, TRIDENT matched the transient and steady-state behavior of these tritium diffusion experiments. The corrosion model in TRIDENT was compared against the natural convection flow-loop experiments at the Oak Ridge National Laboratory (ORNL) from the 1960s and early 19705 which used Molten Salt Reactor Experiment (MSRE) fuel-salt containing UF4. Despite the lack of data required by TRIDENT for modeling the loops, some reasonable results were obtained. The TRIDENT corrosion rates follow the experimentally observed dependence on the square root of the product of the chromium solid-state diffusion coefficient with time. Additionally the TRIDENT model predicts mass transfer of corrosion products from the hot to the cold leg (as was observed in the experiments with salts containing UF4). In a separate paper the results of TRIDENT simulations in a prototypical FHR are presented. (C) 2016 Elsevier B.V. All rights reserved.
机译：氟化物盐冷却高温反应堆（FHR）是一种卵石床核反应堆概念，由液态氟化物盐冷却，称为“氟化物”（（LiF）-Li-7-BeF2）。已开发了可与FHR一起使用的TRITium扩散演化和传输模型（TRIDENT），并以实验数据为基准，TRIDENT是第一个通过中子trans变将盐产生的salt产生的影响与化学氧化还原势的影响相结合的模型，tri传质，tri在管壁中的扩散，graphite对石墨的吸收，氟化tri对铬的选择性腐蚀以及腐蚀产物的传质。尽管不存在来自包含tri的熔融盐的强制对流多热回路的数据进行比较，但将TRIDENT计算与来自游离盐和flinak（0.465LiF-0.115NaF-0.42KF）盐的静态盐扩散测试数据进行了比较。在每种情况下，TRIDENT都与这些tri扩散实验的瞬态和稳态行为相匹配。将TRIDENT中的腐蚀模型与1960年代至19705年代在橡树岭国家实验室（ORNL）的自然对流流动环实验进行了比较，该实验使用了含UF4的熔融盐反应堆实验（MSRE）燃料盐。尽管缺少TRIDENT进行环路建模所需的数据，但仍获得了一些合理的结果。 TRIDENT腐蚀速率遵循实验观察到的对铬固态扩散系数随时间乘积的平方根的依赖关系。此外，TRIDENT模型可预测腐蚀产物从热态到冷态的质量转移（在含UF4盐的实验中已观察到）。在另一篇论文中，介绍了原型FHR中TRIDENT仿真的结果。 （C）2016 Elsevier B.V.保留所有权利。

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