Analysis and Research of Neutron Spectrum in Reactor Shield Calculation

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. (authors)