Neutron-Photon Source Generation Algorithm For PWR Shielding Calculations

The aging and deterioration degree of the main shielding structure material affects the operating life of the reactor,and accurate calculation of the irradiation embrittlement level of the material can provide data support for extending the life of nuclear power plants.The discrete ordinates method,as one of the deterministic shielding calculation methods,can achieve the spatial distribution calculation of neutron-photon flux density inside the reactor,and then obtain the radiation damage assessment parameters of materials.The apparent differences between neutronphoton source are related to their geometric model,burnup,and power distribution,and the numerical accuracy seriously affects the accuracy and reliability of the transport results.Based on the core model structure of pressurized water reactor,this paper proposes a multi-weight source mesh mapping method for neutron-photon source calculation,and the cross section average method is used to weighted average the calculation parameters,the neutron importance is used as the weight to realize the mesh source assignment mapping,reduce the impact of geometric modeling,burnup and power distribution,and improve the accuracy of mesh source calculation.The program can be applied to the source assignment of standard structured meshes and non-uniform discontinuous meshes.Analyze the differences in the distribution of fixed source energy spectra under different evaluation nuclear databases,and evaluate the impact of calculating fixed source energy spectra on the accuracy of transport calculations.The materials of different devices in the reactor are continuously irradiated by high-energy neutrons,which generate radioactive nuclides through activation and release a large number of photons through decay.Based on the transport-activation-transport coupling interface program,the photon flux density distribution in the reactor is calculated,and the photon dose equivalent rate distribution of each structural in the reactor is analyzed,so as to achieve scientific and accurate protection and ensure the safety of workers.The numerical results show that the source calculation of typical pressurized water reactor core model using multi-weight source mesh mapping method can more accurately realize the source assignment of geometric mesh,avoid complex and redundant geometric modeling of the core,and improve the calculation efficiency.NUREG/CR-6115 benchmark question indicates that the calculation accuracy of fast neutron fluence rate in pressure vessels using fixed source generation modules is 20%higher than that of uniform approximate calculations,and it is verified that the main contribution of radioactive activity generation rate in the cavity is 41Ar,and the activated photon source of the biological shielding layer contributes the most to the dose equivalent rate of the outer wall.H.B.Robinson-2 benchmark question uses a fixed source generation module to verify the impact of the evaluation kernel database on the transport calculation results.The program can improve the accuracy of non-uniform discontinuous mesh source assignment.The research in this article helps to generate accurate neutron-photon source distribution,improve shielding transport calculation accuracy,and obtain reactor life extension evaluation parameters.