### About this course

### What you'll learn

- Define basic processes that may occur in the reactor core, laws, equations, and the limits of applicability of models describing the neutron field in the reactor;
- Demonstrate practical experience of calculating the distribution of neutrons in media;
- Demonstrate the ability to analyze the process of slowing down neutrons in various media (typical for nuclear fission reactors) from the standpoint of understanding the physics of the process;
- Evaluate important reactor parameters including performance and safety.

**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

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