Abstract:- This paper describes a numerical procedure to solve the homogeneous boundary problem for a stationary transport equation. The stability and convergence of the proposed finite differences scheme is proved. The error value that corresponds to the numerical solution is also obtained using the Lax theorem. Key-Words: integral-differential equation, finite differences scheme, integral identity method.
172 p.Thesis (Ph.D.)--University of Illinois at Urbana-Champaign, 1980.The multigroup neutron diffus...
Abstract: The numerical algorithm is developed for solving the multigroup steady-state tra...
The paper deals with general properties of the odd-order SP N equations. The equivalence between the...
A general numerical procedure, called the discrete S/sub n/ method, for solving the neutron transpor...
Abstract:- An algorithm for determining the solution of a boundary value problem for an integral-dif...
This paper presents an iterative method based on a self‐adjoint and m‐accretive splitting for t...
We study the spatial discretization, in a fully discrete scheme, for the numerical solution of a mod...
A new numerical method, the Boundary and Interface Function (BIF) Method, for solving the one-dimens...
We study the spatialdiscretization for the numerical solution of a model problem for theneutron tran...
Includes bibliographical references (leaves 265-277)A new numerical method is developed to solve the...
Finite difference techniques are used to solve a variety of differential equations. For the neutron ...
A numerical solution of the first-order, mono-energetic neutron transport equation is found by the f...
Computer codes involving neutron transport theory for nuclear engineering applications always requir...
We prove a regularity result for a Fredholm integral equation with weakly singular kernel, arising i...
Ph.D.Applied SciencesEnergyNuclear engineeringUniversity of Michigan, Horace H. Rackham School of Gr...
172 p.Thesis (Ph.D.)--University of Illinois at Urbana-Champaign, 1980.The multigroup neutron diffus...
Abstract: The numerical algorithm is developed for solving the multigroup steady-state tra...
The paper deals with general properties of the odd-order SP N equations. The equivalence between the...
A general numerical procedure, called the discrete S/sub n/ method, for solving the neutron transpor...
Abstract:- An algorithm for determining the solution of a boundary value problem for an integral-dif...
This paper presents an iterative method based on a self‐adjoint and m‐accretive splitting for t...
We study the spatial discretization, in a fully discrete scheme, for the numerical solution of a mod...
A new numerical method, the Boundary and Interface Function (BIF) Method, for solving the one-dimens...
We study the spatialdiscretization for the numerical solution of a model problem for theneutron tran...
Includes bibliographical references (leaves 265-277)A new numerical method is developed to solve the...
Finite difference techniques are used to solve a variety of differential equations. For the neutron ...
A numerical solution of the first-order, mono-energetic neutron transport equation is found by the f...
Computer codes involving neutron transport theory for nuclear engineering applications always requir...
We prove a regularity result for a Fredholm integral equation with weakly singular kernel, arising i...
Ph.D.Applied SciencesEnergyNuclear engineeringUniversity of Michigan, Horace H. Rackham School of Gr...
172 p.Thesis (Ph.D.)--University of Illinois at Urbana-Champaign, 1980.The multigroup neutron diffus...
Abstract: The numerical algorithm is developed for solving the multigroup steady-state tra...
The paper deals with general properties of the odd-order SP N equations. The equivalence between the...