Abstract
This chapter examines the determinants of neutron multiplication with primary emphasis on the kinetic energy of the neutrons. The chapter begins with discussing the properties of nuclear fuel and of materials that moderate the neutron spectrum. The chapter provides a more detailed description of the energy distributions of neutrons in nuclear reactors, and then discusses the averaging of neutron cross sections over energy. The discussions presented in this chapter maintain the assumption that all neutrons are produced instantaneously at the time of fission. The chapter also discusses about the neutron moderator materials that are required to reduce the neutron energies from the fission to the thermal range with as few collisions as possible. Concepts of neutron energy spectra, the slowing down density, energy self-shielding, thermal neutrons, and energy-averaged reaction rates are also discussed in the chapter. The chapter concludes by defining the neutron multiplication in terms of energy-averaged cross sections.
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