[EN] The spatial distribution of the neutron flux within the core of nuclear reactors is a key factor in nuclear safety. The easiest and fastest way to determine it is by solving the eigenvalue problem of the neutron diffusion equation, which only contains spatial derivatives. The approximation of these derivatives is performed by discretizing the geometry and using numerical methods. In this work, the authors used a finite volume method based on a polynomial expansion of the neutron flux. Once these terms are discretized, a set of matrix equations is obtained, which constitutes the eigenvalue problem. A very effective class of methods for the solution of eigenvalue problems are those based on projection onto a low-dimensional subspace, suc...
Graduation date: 1983Two new concepts have been explored in solving the neutron\ud diffusion equatio...
The Continuous Galerkin Virtual Element Method (CG-VEM) is a recent innovation in spatial discretisa...
In this thesis we study the neutron transport (Boltzmann transport equation) which is used to model ...
Heterogeneous nuclear reactors require numerical methods to solve the neutron diffusion equation (ND...
[EN] Given a configuration of a nuclear reactor core, the neutronic distribution of the power can be...
[EN] Mixed-dual formulations of the finite element method were successfully applied to the neutron d...
summary:We present a method for solving the equations of neutron transport with discretized energeti...
[EN] The use of mixed oxide (MOX) fuel to partially fill the cores of commercial light water reactor...
[EN] The dominant lambda-modes associated with a nuclear reactor configuration describe the neutron ...
In this work we present a methodology of solution of the multigroup multi-layer stationary neutron d...
The solution of the eigenvalue problem for neutron transport is of utmost importance in the field o...
This paper presents new algorithms for use in the eigenvalue response matrix method (ERMM) for react...
We consider high order discontinuous-Galerkin finite element methods for partial differential equati...
A method, called the higher mode synthesis method, for the solution of the two-dimensional neutron d...
To simulate the behaviour of a nuclear power reactor it is necessary to be able to integrate the tim...
Graduation date: 1983Two new concepts have been explored in solving the neutron\ud diffusion equatio...
The Continuous Galerkin Virtual Element Method (CG-VEM) is a recent innovation in spatial discretisa...
In this thesis we study the neutron transport (Boltzmann transport equation) which is used to model ...
Heterogeneous nuclear reactors require numerical methods to solve the neutron diffusion equation (ND...
[EN] Given a configuration of a nuclear reactor core, the neutronic distribution of the power can be...
[EN] Mixed-dual formulations of the finite element method were successfully applied to the neutron d...
summary:We present a method for solving the equations of neutron transport with discretized energeti...
[EN] The use of mixed oxide (MOX) fuel to partially fill the cores of commercial light water reactor...
[EN] The dominant lambda-modes associated with a nuclear reactor configuration describe the neutron ...
In this work we present a methodology of solution of the multigroup multi-layer stationary neutron d...
The solution of the eigenvalue problem for neutron transport is of utmost importance in the field o...
This paper presents new algorithms for use in the eigenvalue response matrix method (ERMM) for react...
We consider high order discontinuous-Galerkin finite element methods for partial differential equati...
A method, called the higher mode synthesis method, for the solution of the two-dimensional neutron d...
To simulate the behaviour of a nuclear power reactor it is necessary to be able to integrate the tim...
Graduation date: 1983Two new concepts have been explored in solving the neutron\ud diffusion equatio...
The Continuous Galerkin Virtual Element Method (CG-VEM) is a recent innovation in spatial discretisa...
In this thesis we study the neutron transport (Boltzmann transport equation) which is used to model ...