Results 271 to 280 of about 5,356 (294)

A Matrix-Free Eigenvalue Solver for the Multigroup Neutron Diffusion Equation

2019
The stationary neutron transport equation describes the neutron population and thus, the generated heat, inside a nuclear reactor core. Obtaining the solution of this equation requires to solve a generalized eigenvalue problem efficiently. The majority of the eigenvalue solvers use the factorization of the system matrices to construct preconditioners ...
Amanda Carreño, Antoni Vidal-Ferràndiz, Damián Ginestar, Gumersindo Verdú   +3 more
openaire   +1 more source

Neutron Diffusion—The Finite Difference Solution of the Neutron Diffusion Equation

2023
openaire   +1 more source

The neutron diffusion equation

2022
openaire   +1 more source

Nonlocal Effects to Neutron Diffusion Equation in a Nuclear Reactor

Journal of Computational and Theoretical Transport, 2020
Rami Ahmad El-Nabulsi
exaly  

Multigroup Neutron Diffusion Equations

2016
openaire   +1 more source

An efficient numerical method for fractional neutron diffusion equation in the presence of different types of reactivities

Annals of Nuclear Energy, 2021
Pradip Roul, Gilberto Espinosa-Paredes, Vikas Rohil   +2 more
exaly  

Modal methods for the neutron diffusion equation using different spatial modes

Progress in Nuclear Energy, 2019
Gumersindo Verdú, A Vidal-Ferràndiz, Damian Ginestar   +2 more
exaly  

Development of a 2D-Multigroup Code (NFDE-2D) based on the neutron spatial-fractional diffusion equation

Applied Mathematical Modelling, 2015
Gilberto Espinosa-Paredes, Hossein Afarideh   +1 more
exaly  

Fractional neutron diffusion equation in space and time

2021
openaire   +1 more source

Design and analysis of a numerical method for fractional neutron diffusion equation with delayed neutrons

Applied Numerical Mathematics, 2020
Pradip Roul, Gilberto Espinosa-Paredes, Vikas Rohil   +2 more
exaly