In this dissertation we present advances to the nonlinear diffusion acceleration for void regions using second order forms of the transport equation. We consider the weighted least-squares and the self-adjoint angular flux transport equations. We show that these two equations are closely related through the definition of the weight function and how the nonlinear diffusion acceleration can be extended to hold in void regions. Using a Fourier analysis we show the convergence properties of our method for homogeneous and heterogeneous problems. We use several problems to study the numerical behavior and the influence of different discretization schemes. Second order forms of the transport equation allow the use of continuous finite elements (...
In this paper we construct numerical schemes to approximate linear transport equations with slab geo...
An Adaptive Mesh Refinement (AMR) technique is presented for the one-group and the multigroup SN tr...
In reactor physics, the efficient solution of the multigroup neutron diffusion eigenvalue problem is...
In this dissertation we present advances to the nonlinear diffusion acceleration for void regions us...
Nuclear engineers are interested in solutions of the Neutron Transport Equation (NTE), with the goal...
In this dissertation, we focus on solving the linear Boltzmann equation -- or transport equation -- ...
Linear discontinuous (LD) spatial discretization of the transport operator can generate negative an...
A simple Richardson iteration procedure converges slowly when applied to thick, diffusive problems w...
In this dissertation, advanced numerical methods for highly forward peaked scattering deterministic ...
We consider the numerical solution of the single-group, steady state, isotropic transport equation. ...
International audienceA new space-angle multi-grid technique has been developed to accelerate the fr...
International audienceWe devise and analyze a new stabilized finite element method to solve the firs...
The diffusion synthetic acceleration (DSA) method has emerged as a powerful tool for accelerating th...
The basic problem addressed in the project was that of accelerating the iterative convergence of Dis...
The linear discontinuous finite element method (LDFEM) is the current work horse of the radiation tr...
In this paper we construct numerical schemes to approximate linear transport equations with slab geo...
An Adaptive Mesh Refinement (AMR) technique is presented for the one-group and the multigroup SN tr...
In reactor physics, the efficient solution of the multigroup neutron diffusion eigenvalue problem is...
In this dissertation we present advances to the nonlinear diffusion acceleration for void regions us...
Nuclear engineers are interested in solutions of the Neutron Transport Equation (NTE), with the goal...
In this dissertation, we focus on solving the linear Boltzmann equation -- or transport equation -- ...
Linear discontinuous (LD) spatial discretization of the transport operator can generate negative an...
A simple Richardson iteration procedure converges slowly when applied to thick, diffusive problems w...
In this dissertation, advanced numerical methods for highly forward peaked scattering deterministic ...
We consider the numerical solution of the single-group, steady state, isotropic transport equation. ...
International audienceA new space-angle multi-grid technique has been developed to accelerate the fr...
International audienceWe devise and analyze a new stabilized finite element method to solve the firs...
The diffusion synthetic acceleration (DSA) method has emerged as a powerful tool for accelerating th...
The basic problem addressed in the project was that of accelerating the iterative convergence of Dis...
The linear discontinuous finite element method (LDFEM) is the current work horse of the radiation tr...
In this paper we construct numerical schemes to approximate linear transport equations with slab geo...
An Adaptive Mesh Refinement (AMR) technique is presented for the one-group and the multigroup SN tr...
In reactor physics, the efficient solution of the multigroup neutron diffusion eigenvalue problem is...