The nine Reactor Statics Modules are designed to introduce students to the use of numerical methods and digital computers for calculation of neutron flux distributions in space and energy which are needed to calculate criticality, power distribution, and fuel burn-up for both slow neutron and fast neutron fission reactors. The diffusion approximation is used for the calculation of neutron transport. Collision probabilities are used to calculate the effect of heterogeneous lattices of fuel and moderator on resonance absorption and slow neutron disadvantage factors. In this module, the three major problems of neutron diffusion in slow neutron fission reactors are reviewed. An energy-dependent diffusion equation, steady state epithermal multi-group equations, and a slow neutron diffusion equation are introduced. The structure and organization of the remaining eight modules are outlined. These modules are intended to supplement textbooks and other lecture material generally available to students in their course work. It is assumed that students are familiar with elementary nuclear structure, neutron-nuclei interactions, and introductory material on fission chain reactors. (Author)