Method for calculating core neutron flux distribution of small experimental reactor

Abstract

Disclosed is a method for calculating core neutron flux distribution of a small experimental reactor. The method includes steps of 1), determining geometrical and material parameters according to a core structure of the small experimental reactor, and establishing a neutron-transport equation describing movement rules of neutrons in each discrete direction inside a reactor core; 2), preparing corresponding quadrature sets for boundaries unparallel in normal surface vectors and each coordinate axis direction; 3), subjecting the neutron-transport equation to numerical discretion to acquire simultaneous linear algebraic equations of angular neutron flux density by a segment method of approximating unstructured geometry with unstructured grids to establish arbitrary triangular grids; 4), solving the simultaneous linear algebraic equations to acquire the discrete distribution of the angular neutron flux density in the reactor core, and acquiring the discrete distribution of neutron-flux density in the reactor core by the aid of relation of the angular neutron flux density and the neutron-flux density. By the method, the neutron-flux distribution of the small experimental reactor, especially that of an isotope production reactor, a pebble bed reactor and a high-flux reactor in the medical field can be accurately acquired.

Description

technical field [0001] The invention relates to the technical field of small experimental nuclear reactors, in particular to a method for calculating the neutron flux distribution in the core of a small experimental reactor. Background technique [0002] Compared with large commercial reactors such as pressurized water reactors, boiling water reactors, heavy water reactors, and fast reactors, small experimental nuclear reactors, such as medical isotope production reactors, pebble bed reactors, and high-flux reactors, generally have complex core geometry, compact structure, and materials. The characteristics of strong non-uniformity and strong anisotropy of neutron angular flux density distribution along the angle make its requirements for the calculation of neutron flux distribution completely different from those of large nuclear reactors. [0003] However, the existing methods for calculating the neutron flux distribution in the core of nuclear reactors are mainly aimed at...

AI Technical Summary

Problems solved by technology

[0005] In order to solve the problems existing in the prior art, the object of the present invention is to provide a method for calculating the neutron flux distribution in the core of a small experimental nuclear reactor, which can accurately obtain the distribution of the neutron flux in the core of a small experimental nuclear reactor, especially to obtain For the neutron flux distribution of the three reactor types of medical isotope production reactor, pebble bed reactor and high flux reactor, the present invention minimizes the approximation brought about by geometric grid division, reduces the numerical discrete process and the iterative calculation process The introduced approximation gives high-precision neutron flux distribution in the core of a small experimental nuclear reactor, which makes up for the shortcomings of the existing technology that cannot give results that meet engineering precision

Images