Neutron Flux and Importance Distribution by Collision Method, Starting from a Generalized Source

In this paper a multigroup method in finite monodimensional geometry is presented for neutron distribution calculation from a given source distribution. This method is extended to neutron importance function calculation for a given detector distribution. Furthermore, it is shown how truncating the calculation at whatever collision, it is possible to evaluate the “residual” neutron distribution by means of usual methods in diffusion or transport theory. An application is presented related to the generalized perturbation methods and the numerical solution of problems of general interest.