氚作为聚变燃料，在自然界丰度极低，国际热核聚变实验堆（ITER）的运行将消耗大量的商业用氚，因此，未来聚变电站必须依靠聚变堆包层部件增殖足够的氚以维持聚变堆的自持运行。我国主导的中国聚变工程实验堆（CFETR）的科学目标之一是验证氚自持，现已进入工程设计阶段。氦冷固态包层作为其首选的包层设计及氚工程提氚系统设计，是聚变堆能否实现氚自持运行的关键。本文以核工业西南物理研究院（SWIP）提出的一种氦冷固态包层设计为研究对象，评估其产氚性能能否满足聚变堆氚自持需求，并进行氚输运行为分析，为提氚系统设计提供参考。首先基于平均氚滞留时间法，对CFETR新的氚循环流程进行稳态运行模式分析，研究快循环模式的引入及其他系统运行参数对聚变堆氚自持性的影响。并对SWIP提出的氦冷固态包层概念进行中子学计算并进行产氚优化设计，评估其氚增殖性能。为获得氦冷包层氚输运行为分析所需源项，对氚增殖材料正硅酸锂进行释氚动力学分析。在原有的界面层+水层释氚模型基础上，给出物理吸附水及化学吸附水吸附上限，并添加化学吸附水解吸项。针对该释氚模型的局限性，提出了界面层+OT层释氚模型，新的释氚模型可以解释不同载气组分下正硅酸锂的释氚行为。通过释氚动力学分析指出了正硅酸锂释氚速率及释氚形式的影响因素。通过DEM+CFD分析方法，开展正硅酸锂球床传质传热特性研究，获得了用于多孔介质模型氚输运分析的宏观等效参数。通过CFD计算获得吹扫气体流动阻力特性，流动压降可由Blake-Kozeny方程计算，拟合得到Blake-Kozeny方程中粘性阻力系数。受球床结构影响，氚在吹扫气体内的等效扩散系数减小。内热源工况下球床等效导热系数低于外加热流工况下的球床等效导热系数。通过流动传热计算验证了吹扫气体与正硅酸锂球床间换热量极低，由多孔介质模型开展包层氚输运分析时可采用局部热平衡模型。针对典型氦冷包层模块，开展包层氚输运行为分析，获得了氚在包层结构材料内的滞留量与向冷却剂的渗透量。并开发了中子、热工、氚输运计算框架，可用于极向各包层模块的氚输运行为分析。
Tritium, a fusion fuel, is in extremely low abundance in nature and the operation of the International Thermonuclear Experimental Reactor (ITER) will consume large amounts of commercially available tritium. Therefore, future fusion power plants must rely on blankets of fusion reactors to proliferate enough tritium to maintain self-sustaining operation of the fusion reactor. The China Fusion Engineering Test Reactor (CFETR) initiated by China has entered the engineering design phase, and the helium-cooled ceramic breeder blanket, which is the preferred blanket design solution and tritium plant tritium extraction system design, is the key factor of whether the fusion reactor can achieve tritium self-sustaining operation. In this paper, a helium-cooled ceramic breeder blanket design proposed by the Southwest Institute of Physics (SWIP) is investigated to evaluate whether the tritium production performance can meet the tritium self-sustainability requirement of fusion reactor, and tritium transport analysis is carried out to provide a reference for the tritium extraction system design. Firstly, a steady-state operation mode analysis of the new tritium cycling process of CFETR is performed based on the mean tritium residence time method. The effects of the introduction of the fast-cycle mode and other system operating parameters on the tritium self-sustainability of the fusion reactor are investigated. Neutronics calculations and optimization for the helium-cooled ceramic breeder blanket concept proposed by SWIP are also performed to evaluate its tritium breeding performance. In order to obtain the source terms required for the analysis of tritium transport behavior in the helium cooled blanket, the tritium release kinetics of lithium orthosilicate, a tritium breeding material, is analyzed. Based on the original model of tritium release from the interface layer + water layer, the upper limits of physisorbed water and chemisorbed water adsorption are given. The desorption term of chemisorbed water is also added. In view of the limitations of this model, the tritium release model of interfacial layer + OT layer is proposed. The new tritium release model can explain the tritium release behavior of lithium orthosilicate under different sweep gas conditions. The factors influencing the rate and form of tritium release from lithium orthosilicate are pointed out through tritium release kinetic analysis. the study of mass and heat transfer characteristics of lithium orthosilicate pebble bed is carried out by the means of DEM+CFD analysis method. Macroscopic equivalent parameters for tritium transport analysis in porous media model are obtained. The calculation results show that the complex and tortuous internal structure of the pebble bed causes an increase in the average transport path of gas molecules, the diffusion coefficient of tritium gas within the purge gas is reduced; The pressure drop of the purge gas inside the pebble bed can be determined by Blake-Kozeny equation, the viscous drag coefficient is obtained by data fitting. The equivalent thermal conductivity of the pebble bed in the internal heat source condition is lower than that of the external heat flux condition. The analysis of tritium transport behavior in the helium cooled blanket is carried out for a typical blanket module. The tritium retention within blanket structure material and tritium penetration into the coolant are obtained. A framework for neutron, thermal and tritium transport calculations has been developed. This framework can be used to analyze the tritium transport behavior of each helium cooled blanket module in the polar direction.
