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High-accuracy fast neutron reactor assembly few-group cross section obtaining method

A high-precision, fast neutron technology, applied in the field of nuclear reactor core design and nuclear reactor physical calculation, can solve problems such as unsatisfactory fast neutron reactor calculation

Inactive Publication Date: 2016-11-16
XI AN JIAOTONG UNIV
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Problems solved by technology

However, due to the great difference in neutron characteristics between fast reactors and thermal reactors, the methods used by these programs are not suitable for the calculation of fast neutron reactors. Therefore, it is necessary to develop a fast neutron reactor with high precision section method

Method used

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  • High-accuracy fast neutron reactor assembly few-group cross section obtaining method
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  • High-accuracy fast neutron reactor assembly few-group cross section obtaining method

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Embodiment Construction

[0067] Below in conjunction with accompanying drawing and specific embodiment the present invention is described in further detail:

[0068]In the present invention, by directly using the point section file, carefully considering the resonance effect existing in the fast reactor components, solving the neutron transport equation of the uniform problem, obtaining the real neutron flux distribution of the components, and then merging the energy groups to obtain Exact minority group section parameters, such as figure 1 Shown, this invention comprises the steps:

[0069] Step 1: For any fast reactor component that needs to be calculated, read the information in the multi-group database, including the total cross section, inelastic scattering cross section, neutron fission spectrum, and the number of neutrons released by each fission, and then firstly calculate the temperature of the cross section Interpolation, where the temperature interpolation of the section follows the law, a...

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Abstract

Provided is a high-accuracy fast neutron reactor assembly few-group cross section obtaining method. The fine-group cross section of a fast neutron reactor assembly is calculated in the mode that point cross section data and multi-group data are combined, online calculation of an elastic scattering matrix is performed, a neutron transport equation is solved by using the fine-group cross section to obtain fine-group neutron-flux density, energy groups are merged on the basis, and accordingly few-group parameters of the fast reactor assembly are obtained. Due to the fact that the method directly uses the point cross section data, the elastic scattering resonance effect of a medium mass nuclide and the resonance interference effects of all nuclides are accurately considered. The neutron-flux density of a real system is adopted in the energy group merging process, and merging results are more accurate. The overall errors of fast reactor assembly few-group parameters calculated by adopting the method are within 1% compared with a reference value, and the method has higher accuracy.

Description

technical field [0001] The invention relates to the fields of nuclear reactor core design and nuclear reactor physical calculation, in particular to a method for obtaining high-precision fast neutron reactor component few-group section. Background technique [0002] At present, the development of fast reactors is in the pre-design and experimental stage. In order to quickly and accurately solve the neutronics parameters of the core, the method based on the "two-step method" has become the main method for fast reactor engineering calculations. The so-called "two-step method", the first step is to carry out multi-group neutron transport calculations for various component materials in the core, obtain the neutron flux density distribution in the components, and then merge the few-group homogenization group cross-section; the second The first step is to solve the minority group neutron transport equation for the core with the homogenization parameters calculated in the previous ...

Claims

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Application Information

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Patent Type & Authority Applications(China)
IPC IPC(8): G21C5/00G21C1/02
CPCG21C1/02G21C5/00Y02E30/30
Inventor 郑友琦杜夏楠曹良志李云召吴宏春
Owner XI AN JIAOTONG UNIV
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