On a boundary-value problem posed by cancer therapy with neutron beams

Abstract

AbstractThis paper contains a detailed formulation of advanced tumor therapy with neutron beams as a mixed boundary initial value problem for multigroup neutron diffusion in a composite 3D multiregional system. By applying a vector-matrix composite region finite-integral transformation we derive the principal operational solution to this problem as the group-regional neutron flux distribution inside the tumor 3D subregions. The principal solution is then converted into expressions of various order approximation, which may be directly programmed on a computer.