Abstract Advanced reactor applications that use a molten fluoride salt coolant and graphite moderator are under consideration as next-generation energy technologies. For molten salts with lithium or beryllium, such as flibe (2LiF-BeF2), the production of tritium from neutron irradiation is a significant technical challenge. Understanding the expected quantities and mechanisms for tritium retention in graphite is important for designing tritium management strategies in these advanced reactors. In this work, the tritium content of IG-110U graphite from a 2013 in-core flibe irradiation experiment was measured by leaching in water and thermal desorption. Five total samples were tested, with an average measured tritium content per salt-contacting surface area of 3.83 ± 0.25 Ci/m2. The tritium measured from the thermal desorption experiments was primarily in a water-insoluble form. Compared to the overall tritium generation during the irradiation, the total amount of retention in graphite predicted by the desorption measurements is significant.

中文翻译：

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熔融氟化盐辐照产生的核石墨中的氚含量和化学形态

摘要 使用熔融氟化盐冷却剂和石墨慢化剂的先进反应堆应用正在考虑作为下一代能源技术。对于含锂或铍的熔盐，如 flibe (2LiF-BeF2)，中子辐照生产氚是一项重大的技术挑战。了解氚保留在石墨中的预期数量和机制对于设计这些先进反应堆中的氚管理策略很重要。在这项工作中，通过在水中浸出和热解吸来测量来自 2013 年芯内纤维辐照实验的 IG-110U 石墨的氚含量。总共测试了五个样品，每个盐接触表面积的平均氚含量为 3.83 ± 0.25 Ci/m2。从热解吸实验中测得的氚主要是不溶于水的形式。与辐照期间产生的总氚相比，解吸测量预测的石墨中总保留量是显着的。