A Calculation Method of Absolute Neutron Flux Spectrum of Coolant in Core Active Zone

Abstract

The invention relates to a computing method of a neutron flux spectrum in a reactor, particularly to a computing method of an absolute neutron flux spectrum of a refrigerant in a reactor core active section. According to the computing method, a method combining reactor core neutron transport program with Monte Carlo program is used, the reactor core computing program is used for computing the absolute value of average flux of the reactor core, the flux ratio of the refrigerant and the reactor core, and the neutron flux distribution of a customized energy group in the refrigerant are generated through analog computation of the Monte Carlo program, and the absolute neutron flux spectrum distributed in different energy groups of the refrigerant in the reactor core active section is finally computed, so that the computing method of traditionally using the value of average neutron flux to compute related source items of the reactor is improved, and the computing accuracy is improved.

Description

technical field [0001] The invention relates to a calculation method for a neutron flux spectrum of a reactor, in particular to a calculation method for an absolute neutron flux spectrum of a coolant in an active zone of a reactor core. Background technique [0002] A nuclear reactor is a device that releases energy through neutron chain fission reactions. It is a relatively large and complex system. In order to make the reactor operate safely and economically, it is often necessary to know the distribution of neutron flux in the reactor. The distribution of neutron flux in the reactor during operation is affected by different factors in the reactor, so the calculation of its distribution in time and space is relatively complicated. In order to meet different needs, the distribution of neutron flux in different regions of the reactor can be used It is obtained by different calculation methods, and the calculation accuracy and time consumption obtained by different methods ar...

AI Technical Summary

Problems solved by technology

The commonly used core neutron transport calculation program can quickly give the absolute value of the neutron flux in the reactor under different core states. flux, and often only the distribution of two or few groups is given. In some calculations such as source term calculations, the neutron flux distribution of the few groups is not enough to meet the accuracy requirements

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