中子能谱在反应堆屏蔽计算中的应用分析研究

Analysis and Research of Neutron Spectrum in Reactor Shield Calculation

在反应堆的屏蔽设计中多采用蒙特卡罗中子-光子耦合输运程序(MCNP)计算反应堆压力容器和堆内构件的中子注量率,用以评估中子对结构材料的辐照损伤。MCNP在计算这类固定源问题时,源强的能量分布多采用MCNP自带的Maxwell裂变中子能谱或Watt裂变中子能谱,它们是典型能量的入射中子对应的向量裂变能谱。然而真正的裂变中子能谱是与入射中子能量相关的矩阵裂变中子能谱。为此,不同的中子能谱对反应堆屏蔽设计计算结果的影响被分析。结果表明:在反应堆屏蔽设计中应考虑不同能量的入射中子对裂变中子能谱的影响,即应该采用矩阵裂变中子能谱进行反应堆屏蔽设计计算。

The monte-carlo neutron-photon coupled transport program(MCNP)is used in reactor shielding design to calculate the neutron flux rate of reactor pressure vessel and reactor components to evaluate the neutron irradiation damage to structural materials.In the calculation of such fixed source problems,the energy distribution of the source is mostly based on the Maxwell fission neutron energy spectrum or the Watt fission neutron energy spectrum of MCNP,which are the vector fission energy spectrum corresponding to the incident neutron of typical energy.However,the real fission neutron energy spectrum is the matrix fission neutron energy spectrum related to the incident neutron energy.Therefore,the influence of different neutron energy spectrum on the reactor shielding design is analyzed.The results show that the influence of incident neutrons of different energies on the fission neutron energy spectrum should be considered in reactor shielding design.