This engineering course is designed to introduce students to a range of concepts, ideas and models used in nuclear reactor physics. This course will focus on the physical theory of reactors and methods of experimental studies of the neutron field.
This course is based on the course “Neutron transport theory” which has been taught at the National Research Nuclear University “MEPhI” for the past 20 years.
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.
Introductions to Nuclear Reactor Physics
-In this module students will learn about the neutron field and main functions to describe it. Students will study the concepts of neutron flux, net current, one-way currents and vector of net current.
Neutrons space behaviour by diffusion aproximation
-In this module students will get knowledge concerning neutrons space behavior by diffusion approximation. Students will learn about diffusion theory, diffusion equation and Fick’s Law.
Main Principals of Slowing down of Neutrons
-In this module students will get knowledge concerning Main Principals of Slowing down of Neutrons. Students will learn about Neutron Spectrum in Absorbing and Non-Absorbing Medium.
Neutron Spectrum in Non-Absorbing and Absorbing Medium
-In this module we will find the dependency of neutron flux by energy (the neutron spectrum). After studding this module, the student should be able to understand and explain the terms in slowing down equation.
Thermalization of Neutrons and MultiGroup Method
-In this module students will learn about main principles of neutron behavior in thermal range. Students will study the ideas of find thermal neutron flux — Maxwell’s spectrum.