【摘要】 针对热管式空间反应堆,基于Open MC程序产生均匀化截面参数,并由确定论快堆分析程序SARAX进行堆芯输运及燃耗计算。以蒙特卡罗程序(MCNP)的输运计算结果以及MVP程序的燃耗计算结果作为参考解,通过对比稳态输运计算和燃耗计算的结果,证明了耦合的Open MC和SARAX程序系统对于空间堆中子学分析和燃耗分析的适用性和高效性。为热管式空间反应堆的设计分析提供了参考。更多还原

【Abstract】 For the heat pipe cooled space reactor, region-dependent homogenized cross sections in the predefined 26 group structure were generated with the OpenMC code based on the R-Z geometric model of the reactor core. The neutron transport calculation was performed with SARAX, which was a deterministic neutronic analysis code for fast spectrum reactors. The calculation results were compared with those obtained with MVP. The generation procedure of the homogenized cross sections was verified and the capability of SARAX for the neutronic analysis of heat pipe reactors was demonstrated.更多还原