Improved method for calculating three-dimensional neutron flux density fine distribution of reactor core

Abstract

The invention discloses an improved method for calculating three-dimensional neutron flux density fine distribution of a reactor core. The method comprises the following steps that 1, modeling is conducted on the reactor core, needed parameters are calculated, and variables are initialized; 2, a radial leakage item is unfolded in the axial direction through a Legendre polynomial expansion, a one-dimensional discrete ordinate difference equation is calculated according to information obtained in the step 1, and an axial leakage item is acquired; 3, two-dimensional spatial distribution of the axial leakage item is obtained through calculation, two-dimensional transportation calculation is conducted according to the information obtained in the step 1, and neutron flux density distribution, a cell-homogenized cross section and the radial leakage item are acquired; 4, whether a characteristic value and the three-dimensional neutron flux density are convergent or not is judged, if not, iteration is continuously conducted from the step to the step 2 till the problems are convergent, and then three-dimensional neutron flux density fine distribution is obtained. According to the method, Legendre polynomial expansion unfolding is conducted on the radial leakage item in the axial direction, the calculation precision is improved, and two-dimensional spatial distribution of the axial leakage item is calculated, so that calculation is closer to true conditions.

Description

technical field [0001] The invention relates to the fields of nuclear reactor core design and reactor physical calculation, in particular to a method for improving the fine distribution of three-dimensional neutron flux density in the reactor core. Background technique [0002] The design and operation of the reactor core need to accurately and quickly calculate the three-dimensional neutron flux density distribution in the reactor and related equipment. The traditional reactor physical analysis and calculation methods widely used at present cannot meet the engineering calculation requirements gradually as the reactor core design becomes more and more complex and the safety requirements become higher and higher. [0003] The so-called traditional reactor physical analysis and calculation method is also called "two-step method". The first step is to perform multi-group neutron transport calculations on various non-uniform components under total reflection boundary conditions,...

AI Technical Summary

Problems solved by technology

However, the handling of leakage items in this method is somewhat simple at present, and some improvements need to be made

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