Multigroup discretisationΒΆ
If we take the standard neutron transport equation:
and integrate it over energy using the following assumptions:
Cross sections are assumed to be constant and independent of the solution.
Angular dependence of total cross section is ignored (arising from integrating over a angularly dependent angular flux).
The infinite sum over scattering order \(\ell\) is truncated to a finite maximum value \(L\). Even for highly anisotropic scattering (important in fusion due to the high energy neutrons), relatively low values are acceptable: the standard FENDL fusion nuclear data library provides \(L=6\).
it can be shown (see here for a detailed derivation) that the resulting equation is:
where \(\psi_{g}(\mathbf{r}, \Omega) = \int_{E_g}^{E_{g-1}} \psi(\mathbf{r}, \Omega, E) dE\) is the group-integrated angular flux, \(Q_g(\mathbf{r}, \Omega) = \int_{E_g}^{E_{g-1}} Q(\mathbf{r}, \Omega, E) dE\) is the group-integrated external source, \(\sigma_{g}(\mathbf{r}, \Omega)\) is the group-averaged total cross section, \(Y_{\ell}(\Omega)\) are the real spherical harmonics (\(\ell = (l,m)\)), and \(\sigma^s_{\ell, g, g' }(\mathbf{r})\) are the Legendre moments of the group-averaged scattering cross section from \(g'\) to \(g\).
By convention, we take the energy groups such that \(E_0\) is the highest energy and \(E_{G}\) is the lowest energy.
Thus, the multigroup transport equation is a discrete set of coupled spatial-angular PDEs, one for each energy group.