A Computational Method for Coupling Analysis of Neutron Transport Fuelup in Fast Neutron Reactors

Abstract

A calculation method for fast reactor neutron transportation burn-up coupling analysis comprises the steps of 1, dividing a radial structure of a reactor core into a triangular grid, establishing a triangular prism grid, dividing burn-up regions with axial segmentation of a subassembly as a unit, dividing internal cycle of a reactor into a plurality of burn-up steps, and executing the following steps for each burn-up step; 2, calculating a small group of macroscopic cross-sections of each burn-up region, and carrying out neutron transportation calculation by use of a neutron transportation calculation method based on the triangular prism grid; 3, calculating a burn-up matrix of each burn-up region at the beginning of the burn-up step, and solving a burn-up equation by use of a Chebyshev rational approximation method; 4, carrying out neutron transportation calculation on the reactor core according to a nucleon density vector of each burn-up region at the end of the burn-up step; 5, averaging the burn-up matrixes at the beginning and at the end of the burn-up step to obtain an average burn-up matrix of each burn-up region, and carrying out burn-up calculation on each bur-up region again from the beginning of the burn-up step; and 6, repeating steps 4 and 5 until the nucleon density vectors of each burn-up region at the end of the burn-up step which are obtained through two adjacent calculations are converged.

Description

technical field [0001] The invention relates to the field of physical calculation and design analysis of nuclear reactors, and relates to a calculation method for coupling analysis of neutron transport fuel consumption in fast neutron reactors. Background technique [0002] With the continuous development of nuclear energy and the continuous expansion of application requirements, new fast neutron reactor design schemes have been continuously proposed. The interior of the core is no longer a single, regular square or hexagonal component arrangement. The fast neutron reactor The design analysis poses new challenges to the accuracy and efficiency of the core neutron transport burnup coupling analysis method. The existing coupled neutron transport burnup analysis methods for fast neutron reactors are mainly based on hexagonal nodal neutron transport or diffusion calculation methods, and linear chain analytic burnup algorithms or matrix exponential burnup algorithms are used for ...

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Problems solved by technology

[0003] The existing coupling calculation methods for neutron transport and burnup of fast neutron reactors mainly have the following two problems: First, they are only suitable for the regular geometric structure of the core, and cannot handle the new type of fast neutron reactor with irregular core layout. Design analysis; secondly, the neutron transport calculation and the fuel consumption calculation are coupled by the sub-step method or the estimation and correction method. When the fuel consumption step size is large, the calculation accuracy will be lost

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