Skip to content

diffusionSolver

Description

In the diffusionSolver solver, the neutron flux distribution is obtained from the solution of the one-group neutron diffusion equation using fitted parameters.

Formulation

The neutron diffusion solver solves the following equation:

\[ \begin{aligned} D\nabla^2 \Phi = \Sigma_A\Phi \end{aligned} \]

where:

  • \(D\) is the neutron diffusion coefficient
  • \(\Phi\) is the neutron thermal flux
  • \(\Sigma_A\) is the macroscopic absorption cross-section

The obtained thermal neutron flux shape is used by the Lassmann burnup class to compute the intra-pin power distribution. The solver considers the fuel pin to be a strongly absorbing medium for thermal neutrons, and therefore only diffusion and absorption terms appear in the equation. The boundary condition for neutron thermal flux needed by the solver can be selected by the user to a value of choice. Since only the thermal neutron shape is of importance for the Lassmann burnup class, the user could simply select the boundary condition for flux to be a fixedValue to 1.

Options

No options available.

Usage

Here is a code snippet of the solverDict to be used for activating the diffusion neutronics solver class:

neutronicsSolver diffusion;