The numerical solution to the one-group time-dependent neutron transport equation in infinite plane, spherical, and cylindrical geometries is obtained via an expansion in Legendre polynomials. The computation features general anisotropic scattering, isotropic and beamsources, and a power law time-dependent cross-section variation. Results for test problems are compared with previously obtained numerical solutions and with the diffusion approximation.
ASJC Scopus subject areas
- Nuclear Energy and Engineering