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Method for calculating coefficient of any order in neutron transport discrete nodal method

A technology of discrete block method and intermediate coefficients, which is applied in calculation, electrical digital data processing, special data processing applications, etc., and can solve the problems of limiting the accuracy and efficiency of numerical methods

Active Publication Date: 2016-12-07
XI AN JIAOTONG UNIV
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Problems solved by technology

This method of calculating coefficients in the literature limits the order of spatial variable expansion within a block, thereby limiting the accuracy and efficiency of the numerical method

Method used

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  • Method for calculating coefficient of any order in neutron transport discrete nodal method
  • Method for calculating coefficient of any order in neutron transport discrete nodal method
  • Method for calculating coefficient of any order in neutron transport discrete nodal method

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Embodiment Construction

[0073] The present invention will be further described in detail below in conjunction with specific embodiments:

[0074] The method of the present invention is based on the discrete nodal method of neutron transport. In order to obtain the expansion coefficient of any order of the discrete nodal method of neutron transport in the neutron transport problem, under the framework of the discrete nodal method of neutron transport, Computer programming implements the previous step code.

[0075] The most basic relational expressions used in programming are (23), (24), and (25). Observing these three relational expressions, we can see that on the one hand, the intermediate coefficients a, b, and c need to be obtained; Such as And J kn The integral of the product of the exponential function and any order polynomial.

[0076] The intermediate coefficients a, b, c and the size of the segment Δx, the cosine value of the angle between the neutron flight direction and the coordinate axis, an...

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Abstract

The invention discloses a method for calculating a coefficient of any order in a neutron transport discrete nodal method. The method mainly comprises the steps of 1, performing form simplification: performing form simplification on complex source coefficient, flux coefficient and coupling coefficient by introducing an intermediate coefficient; 2, performing integral transformation: transforming an integral, appearing in the source coefficient, the flux coefficient and the coupling coefficient, of an exponential function and Legendre polynomial product into an integral of the exponential function and general polynomial product according to properties of a Legendre polynomial; and 3, performing analytic solving: analytically deriving accurate expressions and recursive relations of an intermediate integral through a mathematic method, substituting an intermediate integral value accurately solved by a computer into expressions of the source coefficient, the flux coefficient and the coupling coefficient after integral transformation, and obtaining accurate values of the source coefficient, the flux coefficient and the coupling coefficient of any order in combination with an intermediate coefficient value as a known condition, wherein the accurate expressions and the recursive relations do not contain complex integral operations and are easily realized by computer programming.

Description

Technical field [0001] The invention relates to the field of numerical simulation of the neutron transport process in nuclear engineering, in particular to a method for calculating the expansion coefficient of any order Legendre polynomial in the discrete nodal method of neutron transport. Background technique [0002] Nuclear reactor is a device that realizes controllable self-sustained neutron fission reaction process. Neutron fission reaction is a process in which neutrons and fissionable material nuclei undergo fission reactions to generate fission fragments, new neutrons and photons, etc., and release energy at the same time. This energy can be used by people, which is the principle of nuclear power plants. However, due to the complexity of the self-sustaining chain fission reaction, and the neutrons and photons released directly from the nuclear fission reaction and the indirect release of fission fragments, the numerical simulation of the self-sustaining chain fission rea...

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Application Information

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Patent Type & Authority Applications(China)
IPC IPC(8): G06F19/00
CPCG16Z99/00
Inventor 吴宏春徐志涛李云召郑友琦
Owner XI AN JIAOTONG UNIV
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