A Numerical Simulation Method of Neutron Transport with Personalized Treatment According to Neutron Energy

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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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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