Nuclear Reactor Physics Basics

Section 1. Properties of Free Neutron and Nuclear Fission

• Describe the properties of free neutrons and it's classification
• Indentify principal nuclear reaction - neutron sources
• Define main properties of nuclear fission

Section 2. Interactions of Neutrons with Matter

• Define main process of neutron interaction with nuclei of medium
• Define microscopic and macroscopic cross sections and mean free path

Section 3. Neutron Field and Main Functions to Describe it

• Define neutron flux, net current, one-way currents and vector of net current
• Calculate the functions in simple cases

Section 4. Diffusion Theory. Diffusion equation and Fick 's Law

• Define main approximations of diffusion theory - Fick's Law and diffusion equation
• Explain each term in diffusion equation

Section 5. Solutions of Diffusion Equation in Different Geometries

• Define initial and boundary conditions to find solution
• Find solutions of diffusion equation in different geometries
• Interpret the solutions from physical meaning

Section 6. Solutions of Diffusion Equation in Multiplying Medium

• Find and interpret the solution in multiplying medium
• Define material and geometry buckling, multiplication factor

Section 7. Main Principals of Slowing down of Neutrons

• Define the reason to slow down of neutrons
• Explain what nuclear reaction is the best for slowing down
• Explain main principals of elastic scattering - post collision energy range, frequency function, mean energy loss per one collision etc.

Section 8. Neutron Spectrum in Non-Absorbing Medium

• Define lethargy of neutrons and it's connection to energy
• Explain the terms in slowing down equation
• Find the solution in non-absorbing medium

Section 9. Neutron Spectrum in Absorbing Medium

• Explain the dependency of absorbing cross section to energy
• Define Doppler effect, slowing down density, resonance escape probability
• Find solution of slowing down equation in absorbing medium

Section 10. Thermalization of Neutrons

• Define main principles of neutron behavior in thermal range
• Explain ideas to find thermal neutron flux - Maxwell's spectrum
• Define averaged cross section, Vescott factors and it's dependency on ambient temperature

Section 11. Multi Group Method

• Define energy group
• Define principles of getting averaged cross section
• Explain the multi group approximation
• Find solutions of group diffusion equations