# Modeling Neutron Transport and Interactions

## Abstract

It is essential to know the spatial and energy distributions of the neutrons in a field in a nuclear fission reactor, D–T (or D–D) fusion reactor, or other nuclear reactors populated with large numbers of neutrons. It is obvious why the spatial distribution should be known, and because neutron reactions vary widely with energy, the energy distribution is also a critical parameter. The neutron energy distribution is often called the neutron spectrum. The neutron distribution satisfies transport equation. It is usually difficult to solve this equation, and often approximated equation so-called diffusion equation is solved instead. In this chapter only overview of transport equation and diffusion equation of neutrons is presented, and methods for solving these equations are presented in the following sections.

## References

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