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Neutron transport numerical simulation method for carrying out personalized treatment according to neutron energy

A technology of neutron energy and numerical simulation, applied in the direction of electrical digital data processing, special data processing applications, design optimization/simulation, etc., can solve the problems that do not have the value of engineering design calculations, large quantities, and time-consuming calculations, and achieve The effect of avoiding sampling simulation process, improving calculation efficiency and reducing calculation amount

Active Publication Date: 2018-11-02
XI AN JIAOTONG UNIV
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Problems solved by technology

Therefore, the Monte Carlo method has high calculation accuracy, but in order to obtain a more accurate calculation, a large number of samples and simulations are required, which makes the calculation of the Monte Carlo method seriously time-consuming, and it does not have the ability to calculate the actual reactor physics problems. The value of engineering design calculations

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  • Neutron transport numerical simulation method for carrying out personalized treatment according to neutron energy
  • Neutron transport numerical simulation method for carrying out personalized treatment according to neutron energy
  • Neutron transport numerical simulation method for carrying out personalized treatment according to neutron energy

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Embodiment Construction

[0070] Below in conjunction with accompanying drawing and specific embodiment the present invention is described in further detail:

[0071] In neutron shielding and protection design calculations and reactor physical numerical simulation calculations, the present invention comprehensively takes into account the advantages of fast calculation speed of the existing characteristic line method and high precision of the calculation of the resonance energy region by the Monte Carlo method, by coupling the two methods , to avoid the existing characteristic line method to calculate the resonance energy region has the disadvantage of approximation and the huge calculation amount of the Monte Carlo method, and provides a solution for neutron shielding and protection design and reactor physical numerical simulation.

[0072] like figure 1 Shown, the specific implementation steps of the present invention are as follows:

[0073] Step 1: For the form of the core neutron transport problem...

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Abstract

The invention provides a neutron transport numerical simulation method for carrying out personalized treatment according to neutron energy. By establishing a same flat source area structure and an energy group structure, a fast neutron energy segment and a thermal neutron energy segment are scanned and calculated by using a characteristic line method; Monte Carlo simulation is used in a resonanceenergy segment, and different energy segments are mutually coupled through a scattering source and a fission source term; and fission source iteration is carried out according to the convergence condition, so as to obtain a characteristic value and flux distribution required by the neutron transport numerical simulation. According to the neutron transport numerical simulation method for carrying out the personalized treatment according to the neutron energy provided by the invention, non-resonance energy segments can be processed through the characteristic line method, and the faster calculation speed in the energy segments can be kept; a Monte Carlo method can be adopted to directly sample and simulate resonance energy segments that cannot be accurately calculated by the characteristic line method; and besides, the resonance energy segments can be processed accurately while shortening the calculation time, and the calculation accuracy is improved.

Description

technical field [0001] The invention relates to the fields of neutron shielding and protection design calculation and reactor physical numerical simulation calculation, in particular to a neutron transport numerical simulation method for individualized processing according to neutron energy. Background technique [0002] In neutron shielding and protection design calculations and reactor physical numerical simulation calculations, after the fuel composition, structural material composition and core layout are given, neutron transport numerical simulation calculations are required to obtain the core Effective multiplication factor and flux distribution. [0003] At present, the numerical simulation of neutron transport is mainly divided into two categories. One is called the deterministic method, including collision probability method, discrete ordinate method, characteristic line method, etc., which mainly rely on different discrete approximation techniques. The continuous ...

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Application Information

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Patent Type & Authority Applications(China)
IPC IPC(8): G06F17/50
CPCG06F30/20
Inventor 李云召郑琪吴宏春曹良志
Owner XI AN JIAOTONG UNIV
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