A three-dimensional quasi-transport acceleration method for the uniform geometry variational nodal method

A technology of variational segmental and neutron transport equations, applied in the field of three-dimensional quasi-transport acceleration, can solve problems such as high calculation costs, and achieve the effects of reduced calculation time, reduced calculation memory, and better calculation memory
CN113673116BActive Publication Date: 2022-03-08SHANGHAI JIAOTONG UNIV
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Patent Information

Authority / Receiving Office
CN · China
Patent Type
Patents(China)
Current Assignee / Owner
SHANGHAI JIAOTONG UNIV
Publication Date
2022-03-08

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Abstract

A three-dimensional quasi-transport acceleration method for the uniform geometric variational nodal method, which eliminates the axial and radial cross-derivative terms in the second-order even-parity neutron transport equation through quasi-transport approximation; considering the In the actual pressurized water reactor, the axial inhomogeneity is relatively weak, and the present invention adopts diffusion approximation to the angular distribution of the odd neutron angular flux density on the axial surface in the Ritz discrete process; , taking advantage of the symmetry of the angular space, reduces the number of angular basis functions for the odd neutron angular flux density on the radial surface. The invention can greatly improve the calculation efficiency and reduce the calculation memory without significantly affecting the calculation accuracy, so as to be used for accurate and efficient neutronics simulation in nuclear reactor design.
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Description

technical field

[0001] The invention relates to a technology in the field of nuclear engineering, in particular to a three-dimensional quasi-transport acceleration method for the uniform geometric variational segmental method. Background technique

[0002] In reactor design, in order to analyze the neutronics performance and safety of the reactor, accurate and efficient neutronics simulation of the reactor is required to obtain the effective multiplication coefficient of the reactor and the neutron flux density distribution in the reactor. The effective multiplication coefficient and neutron flux density are obtained by solving the neutron transport equation. At present, the widely used method for solving the neutron transport equation is the variational segment method, which divides the reactor area into a series of typical segments , by constructing the corresponding response matrix for typical nodes and completing the solution of the response matrix equation, the effectiv...

Claims

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