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HFETR裂变中子转换器设计 被引量:1

Design of Fission Neutron Converter in HFETR
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摘要 为提高高通量工程试验堆(HFETR)局部快中子注量率,裂变中子转换器采用以含7%Mo的高裂变密度UMo合金作为燃料芯体的十字形燃料棒。转换器内62根燃料棒以三角点阵布置于63mm外套管和24mm内套管间,中心区域为20mm的辐照孔道。采用蒙特卡罗计算表明,该转换器内辐照样品的快中子(E>1MeV)注量率可达3.34×1014 cm-2.s-1,较堆芯相同位置不放置转换器时高约40%。在HFETR设计流速和压力下,利用ANSYS/CFX程序分析得到,转换器最大允许功率可达2.4MW,燃料棒芯体最大功率密度为8.007kW/cm3。此时,燃料棒包壳温度为193.6℃,能满足HFETR的热工要求,不会产生流动不稳定。 In order to increase local fast neutron fluence rate in High Flux Engineering Test Reactor(HFETR),the fission neutron converter adopted the crisscross fuel rod whose fuel pellet was made of high fission density alloy UMo with 7%Mo.62 fuel rods in the converter were arranged with triangle dot-matrix between outer tube with diameter of 63 mm and inner tube with diameter of 24 mm.And the converter has an irradiation hole with diameter of 20 mm in the center.The calculation result with MCNP shows that fast neutron(E1 MeV) fluence rate of irradiated samples in the converter can achieve up to 3.34×1014 cm-2·s-1,which is about 40% higher than that in the HFETR core at the same position without converter.On the other hand,under the condition of design flow velocity and pressure,the analysis results with ANSYS/CFX show that the maximum permission power can reach 2.4 MW and the maximum power density of fuel pellet is 8.007 kW/cm3.Here,the cladding temperature of the fuel rod is 193.6 ℃,and the converter can fulfill the requirement of thermal-hydraulic design criteria of HFETR and flow instability will not occur.
出处 《原子能科学技术》 EI CAS CSCD 北大核心 2013年第6期988-991,共4页 Atomic Energy Science and Technology
基金 科技部国际热核聚变实验堆(ITER)专项资助项目(2011GB108010)
关键词 高通量工程试验堆 裂变中子转换器 设计 High Flux Engineering Test Reactor fission neutron converter design
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  • 1HENGSTLER R M, BECK L, BREITKREUTZ H, et al. Physical properties of monolithic U8wt. %-Mo[J]. Journal of Nuclear Materials, 2010, 402(9): 74-80. 被引量:1
  • 2DANIEL I, DANIEL P. A European collabora- tion for designing up-to-date irradiation devices for material and nuclear fuels in material test reactor[C]//Proceeding of the 3rd Meeting of the International Group on Research Reactors (IGORR-Ⅲ). Japan: Naka-Machi, 1993. 被引量:1

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