氢化锆慢化熔盐堆钍铀转换性能初步分析

Preliminary analysis of Th-U conversion performance in a ZrH-moderated molten salt reactor

中子能谱对钍基燃料在熔盐堆中的利用效率及温度反馈系数等安全问题有较大影响,所以对熔盐堆新型慢化剂的研究具有重要意义。本工作基于SCALE6计算程序,对不同几何栅元结构的氢化锆栅元组件在熔盐堆的物理性能进行了研究,分别计算了中子能谱、钍铀转换比、^(233)U浓度、总温度反馈系数以及燃耗等中子物理参量。结果表明,减小六边形栅元对边距或者增加熔盐占栅元体积比可以增加钍铀转换比和改善温度反应性系数;当加入的氢化锆慢化剂体积份额为0.1时就可以将熔盐堆^(233)U初始浓度降低到2.5×10^(-2)以内;氢化锆慢化熔盐堆在超热谱条件下,其^(233)U初装载量和超铀核素产量较小,同时堆芯较为紧凑。

Background： The neutron spectrum plays an important role in thorium-based fuel utilization efficiency and temperature feedback coefficient concerning reactor operation safety, so it is very important to study the new moderator material used in molten salt reactor （MSR）.Purpose： This study aims to analyze thethorium-uranium conversion performance of a ZrH-moderated molten salt reactor and analyze the feasibility of ZrH as moderator in molten salt reactor.Methods： SCALE program is used to calculate neutron spectrum, thorium uranium conversion ratio,233U concentration, total temperature feedback coefficient and burnup calculation with different lattice parameters.Results： The thorium uranium conversion ratio and total temperature feedback coefficient can be improved significantly by reducing lattice size or increasing salt volume ratio; the initial233U concentration for start reactor can be easily controlled under 2.5×10^2 when the volume share of added ZrH is 0.1.Conclusion：Compared to the graphite-moderated MSR, ZrH-moderated MSR reduces initial233U inventory and transuraniums （TRUs） production, and makes its core more compact.