Results 21 to 30 of about 4,725 (242)

Pin-by-pin Calculation of Reactor Core-Based on Quasi-diffusion in Rectangular Grid

Yuanzineng kexue jishu, 2023
Physical calculations of reactor cores are fundamental for reactor core design and nuclear safety analysis. The second generation of core calculation theory and methods based on advanced component homogenization theory and modern coarse grid block ...
YAN Jiangtao;LIU Kun;ZHUANG Kun;WANG Senshan;ZHANG Kaihui;ZHANG Xinxin
doaj  

Time integration of the neutron diffusion equation on hexagonal geometries

Mathematical and Computer Modelling, 2010
To study the behavior of nuclear reactors like the Russian VVER reactors it is necessary to solve the time dependent neutron diffusion equation using a hexagonal mesh. Two methods are proposed to solve this equation. In both methods the spatial part of the equations is discretized using a high order spectral element method, based on assuming that the ...
S. González-Pintor, Damián Ginestar, Gumersindo Verdú   +2 more
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Diffusion synthetic acceleration with the fine mesh rebalance of the subcell balance method with tetrahedral meshes for SN transport calculations

Nuclear Engineering and Technology, 2020
A diffusion synthetic acceleration (DSA) technique for the SN transport equation discretized with the linear discontinuous expansion method with subcell balance (LDEM-SCB) on unstructured tetrahedral meshes is presented.
Habib Muhammad, Ser Gi Hong
doaj   +1 more source

Empirical Equation of Neutron Diffusion System [PDF]

Journal of Nuclear Science and Technology, 1970
An empirical equation between the cross sections and criticality factor was derived by the technique of experimental design and regression analysis. Factorial experiments taking account of interactions between factors was designed with use made of the orthogonal array. Numerical calculation of a two-group, two-core system was performed and an empirical
openaire   +1 more source

Discrete Time Deep Learning Solution Method for Transient Neutron Diffusion Equation

Yuanzineng kexue jishu
Solving the neutron diffusion equation is the key to reactor design and analysis. In engineering applications, the neutron diffusion equation is often characterized by multi-dimensionality and multi-energy groups.
YAO Hemin1, ZHANG Heng1, LIU Dong2, HE Yunling1, HANG Qin1, XIANG Di2
doaj   +1 more source

Sensitivity Analysis of the Galerkin Finite Element Method Neutron Diffusion Solver to the Shape of the Elements

Nuclear Engineering and Technology, 2017
The purpose of the present study is the presentation of the appropriate element and shape function in the solution of the neutron diffusion equation in two-dimensional (2D) geometries.
Seyed Abolfazl Hosseini
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Development and Validation of Three-dimensional Neutronics Code Based on Advanced Nodal Method for Hexagonal-z Geometry

Yuanzineng kexue jishu, 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.
LU Daogang;LYU Siyu;SUI Danting;GUO Jinsong
doaj  

Research on neutron diffusion coupling calculation based on the UDS and UDF functions of FLUENT and its application analysis on fast reactor

四川大学学报. 自然科学版, 2020
With the great improvement of computer performance, analyzing the complex flow and heat transfer phenomenon by coupling CFD and neutronics has attracted lots of attentions nowadays. The study aims to investigate the neutron diffusion coupling calculation
ZHANG Xue-Bei, WANG Chi, CHEN Hong-Li
doaj  

Development of the S3-HACNEM Simulator Program in order to Solving the Forward and Adjoint Neutron Diffusion Equation for Rectangular Geometry Reactor Cores [PDF]

مجله علوم و فنون هسته‌ای
In nuclear reactor calculations, such as burn-up and fuel management, transient analysis, and pin power reconstruction, methods are being developed that are optimal, and are both cost-efficient and time-efficient. In this paper, the discretization of the
A. Kolali,, D. Naghavi Dizaji, N. Vosoughi   +2 more
doaj   +1 more source

A numerical validation between the neutron transport and diffusion theories for a spatial kinetics problem

Brazilian Journal of Radiation Sciences
In this paper, a comparative analysis of numerical results of the neutron transport and diffusion theories for steady-state and transient multigroup problems is presented.
Rodrigo Zanette, Liliane Basso Barichello, Claudio Zen Petersen   +2 more
doaj   +1 more source