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One-step transport calculation method and system based on axial flux expansion

A technology of axial flux and calculation method, which is applied in complex mathematical operations, design optimization/simulation, geometric CAD, etc., and can solve problems such as iterative divergence, large amount of calculation, and difficult application

Pending Publication Date: 2021-11-19
NUCLEAR POWER INSTITUTE OF CHINA
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Problems solved by technology

[0005] The technical problem to be solved by the present invention is the physical analysis of the existing reactor, solving the neutron transport equation to obtain the core reactivity and the fine power distribution of the whole reactor, the three-dimensional characteristic line method adopted has a large amount of calculation and is difficult to apply, and the two-dimensional The one-dimensional coupling method has poor stability and iterative divergence, and both methods are difficult to apply to new advanced reactors. The purpose of the present invention is to provide a one-step transport calculation method and system based on axial flux expansion, which solves the problems in the existing Under the calculation conditions, without increasing the amount of calculation, it can achieve good stability and avoid iterative divergence, so as to solve the problem of one-step transportation calculation of advanced reactors
For reactor design calculations

Method used

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  • One-step transport calculation method and system based on axial flux expansion
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  • One-step transport calculation method and system based on axial flux expansion

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Embodiment 1

[0030] This embodiment 1 is a one-step transport calculation method based on axial flux expansion, which takes into account the calculation efficiency of the two-dimensional / one-dimensional coupling method and the calculation stability of the three-dimensional characteristic line method, and overcomes the introduction of the two-dimensional / one-dimensional coupling method. The iterative divergence problem caused by the leakage item realizes the one-step transportation calculation of the whole core of the advanced reactor under the premise of acceptable calculation efficiency.

[0031] A one-step transport calculation method based on axial flux expansion in this implementation 1 is applied to solve the three-dimensional neutron transport equation. The input data of the three-dimensional neutron transport equation is the nuclear reactor geometric structure, cross-section data, etc., and the reactor characteristic value and three-dimensional neutron flux distribution are obtained ...

Embodiment 2

[0042] This embodiment 2 is based on the embodiment 1, a one-step transport calculation system based on axial flux expansion, the system adopts the one-step transport calculation method based on axial flux expansion in embodiment 1, and implements The system in Example 2 specifically includes:

[0043] Data acquisition module: used to acquire CSG geometry module and section data of advanced reactors;

[0044] Equation establishment module: used to establish three-dimensional neutron transport equation, the input items of three-dimensional neutron transport equation are CSG geometry module and section data of advanced reactor;

[0045] Equation solving module: it is used to expand the flux in the axial direction by using the first-order difference form, and integrate the three-dimensional neutron transport equation layer by layer, so as to solve the three-dimensional neutron transport equation and obtain the eigenvalues ​​of the reactor core and three-dimensional neutron flux ...

Embodiment 3

[0048] Embodiment 3 is based on Embodiment 1. Embodiment 3 provides a device, where the device includes: one or more processors;

[0049] memory for storing one or more programs that, when executed by the one or more processors, cause the one or more processors to perform a one-step transport calculation method based on axial flux expansion .

[0050] Wherein, a one-step transport calculation method based on axial flux expansion is performed according to the method steps in Example 1. No more details here.

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Abstract

The invention discloses a one-step transport calculation method and system based on axial flux expansion. The method comprises the steps: building a three-dimensional neutron transport equation, wherein an input item of the three-dimensional neutron transport equation is advanced reactor data; in the axial direction, carrying out flux expansion in a first-order difference mode, and carrying out layer-by-layer integration on a three-dimensional neutron transport equation; and solving the three-dimensional neutron transport equation to obtain a reactor core characteristic value and three-dimensional neutron flux distribution for reactor design calculation. According to the invention, under existing calculation conditions, the calculation amount is not increased, meanwhile, good stability is achieved, iterative divergence is avoided, and the problem of one-step transport calculation of an advanced reactor is solved.

Description

technical field [0001] The invention relates to the field of nuclear reactor core design and reactor physical numerical calculation, in particular to a one-step transport calculation method and system based on axial flux expansion. Background technique [0002] As the basis of nuclear reactor system analysis and calculation, reactor physics analysis and calculation obtains the reactivity of the core and the fine power distribution of the whole reactor by solving the neutron transport equation. In order to quickly carry out the research and development of advanced nuclear power reactor core, it is necessary to develop advanced high-precision reactor physical design software. In order to simulate complex structural cores, researches on "one-step" reactor physical calculation methods based on accurate physical models and fine geometric modeling are being carried out extensively at home and abroad. [0003] The one-step transport calculation method refers to the neutron transpo...

Claims

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Application Information

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IPC IPC(8): G06F17/11G06F30/10G06F30/20
CPCG06F17/11G06F30/10G06F30/20
Inventor 彭星杰赵晨赵文博张斌卢宗健于颖锐李庆
Owner NUCLEAR POWER INSTITUTE OF CHINA
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