混合堆增殖钍基燃料组件中子学分析

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基金项目:

中央高校基本科研业务费专项资金(09MG11)

国家自然科学基金(10705011,10875042)

长江学者和创新团队发展计划(IRT0720)

先进核燃料循环863项目(2009AA050701)

Neutronics Calculation of Fusion-Fission Hybrid Breeding Thorium Fuel Assembly

School of Nuclear Science and Technology, North China Electric Power University, Beijing 102206, China

摘要: 采用压水堆17×17燃料组件模型,用燃料组件参数计算程序DRAGON分别对混合堆增殖钍燃料组件和全铀组件的中子学特性进行了研究,分析组件的燃料温度系数、慢化剂温度系数及其与燃耗的关系。计算结果表明,混合堆增殖钍燃料组件和全铀组件的中子特性相似,但钍燃料组件中的乏燃料组件中的次锕系核素(MA)的含量明显减少。
- 混合堆增殖钍基燃料 /
- 中子学 /
- 温度系数
Abstract: A preliminary comparative study of the physical properties among 17×17 fuel assembly in PWRs for prototype between uranium assembly and hybrid breeding thorium-based assembly has been investigated respectively using the DRAGON software.The parameters such as fuel temperature coefficient,moderator temperature coefficient and that variation as a function of operation period have been investigated.Results show that the neutron properties of uranium-based assembly and hybrid breeding thorium-based assembly are similitude,but MA mass of hybrid breeding thorium-based assembly is evidently less than those of the uranium assembly.
- Fusion-fission hybrid breeding thorium fuel /
- Neutronics /
- Temperature coefficient