Zobrazeno 1 - 10
of 515
pro vyhledávání: '"Neutron Diffusion Equation"'
Publikováno v:
Nuclear Engineering and Technology, Vol 56, Iss 10, Pp 4207-4218 (2024)
One of the essential requirements for molten salt reactor (MSR) design is methodology for analyzing multi-physics phenomena, such as the behavior of liquid fuel. In the research of Molten Salt Fast Reactors (MSFRs), the Neutron Diffusion Equation (ND
Externí odkaz:
https://doaj.org/article/ac2a2c981e9d4c82abca96129cf6522b
Publikováno v:
Nuclear Engineering and Technology, Vol 56, Iss 8, Pp 2948-2957 (2024)
High-order harmonic techniques can be used to recreate neutron flux distributions in reactor cores using the neutron diffusion equation. However, traditional source iteration and source correction iteration techniques have sluggish convergence rates
Externí odkaz:
https://doaj.org/article/00e73864b42843dc942464cdf0b49b47
Publikováno v:
He jishu, Vol 47, Iss 2, Pp 135-141 (2024)
国防科工局核能开发科研项目（cosSYST等热工水力计算分析软件的自主化研发与示范应用）资助
Externí odkaz:
https://doaj.org/article/8bc2e25bdecd430ca431a192fee4b6a5
Publikováno v:
Nuclear Engineering and Technology, Vol 55, Iss 6, Pp 2172-2194 (2023)
A variational nodal method (VNM) with unstructured-mesh is presented for solving steady-state and dynamic neutron diffusion equations. Orthogonal polynomials are employed for spatial discretization, and the stiffness confinement method (SCM) is imple
Externí odkaz:
https://doaj.org/article/aa23f7e44c2c4a1791427bcbf36c867d
Publikováno v:
Yuanzineng kexue jishu, Vol 57, Iss 3, Pp 565-575 (2023)
As the cornerstone of the development of the computational tool for the coupled neutronics and thermal-hydraulics analysis dedicated to the micro gas-cooled reactor, neutronic diffusion solver was developed in this work based on the generic 3D finite
Externí odkaz:
https://doaj.org/article/819a94fb326b4140b343406faf310093
Akademický článek
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Akademický článek
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Publikováno v:
Yuanzineng kexue jishu, Vol 56, Iss 2, Pp 351-359 (2022)
Hexagonal fuel assemblies are widely used in liquid metal-cooled fast reactors (LMFR). The design and safety analysis of these reactors require three-dimensional full-core coupling calculations of neutron fluxes and currents in the core. After years
Externí odkaz:
https://doaj.org/article/4cb890d7d289464bbdfc2bcac9dbc009
Publikováno v:
Journal of Nuclear Engineering, Vol 2, Iss 4, Pp 533-552 (2021)
In this paper, the artificial neural networks (ANN) based deep learning (DL) techniques were developed to solve the neutron diffusion problems for the continuous neutron flux distribution without domain discretization in advance. Due to its mesh-free
Externí odkaz:
https://doaj.org/article/1484d4b61c604969981417148039f50f
Publikováno v:
Brazilian Journal of Radiation Sciences, Vol 10, Iss 3A (2022)
Many numerical methods are being used to solve the multigroup neutron diffusion equation for different types of nuclear reactors. These methods solve this equation quite accurately and determine the neutron flux and power distribution in the reactor
Externí odkaz:
https://doaj.org/article/b428546be59e41d3a7305483fa7c2f4e