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快中子注量是影响压力容器材料性能的重要指标。在堆芯装有钚铀氧化物混合燃料(MOX燃料),堆芯物理特性发生明显变化时,现有的屏蔽计算软件能否准确预测压力容器所受的快中子注量率值得研究。本研究分别使用MCNP、TORT、SCALE等国际通用的屏蔽计算程序对VENUS-2基准题进行分析比较。研究表明,各软件对含MOX燃料堆芯的中子注量率计算偏差均在合理的范围内,能满足工程设计的需求,MCNP程序的计算精度相对更高。
Fast neutron fluence is an important indicator of the material properties of pressure vessels. Whether the existing shielding calculation software can accurately predict the fast neutron fluence rate of a pressure vessel deserves research when there is a significant change in the physical properties of the core with plutonium-uranium oxide hybrid fuel (MOX fuel) in the core. In this study, we used MCNP, TORT, SCALE and other international common mask calculation programs to analyze and compare VENUS-2 benchmark questions. The results show that the deviations of neutron fluence rates of MOX fuel cores are within a reasonable range for each software, which can meet the needs of engineering design. The calculation accuracy of MCNP program is relatively higher.