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    6 Weeks
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    2–3 hours per week
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Acquisition in math and general physics, basics of integration, first and second order differential equations, vector algebra, math analysis, basics of atomic physics.

About this course

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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 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

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  • 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

Meet your instructors

Yury N. Volkov
Senior lecturer, PhD in Nuclear Engineering
National Research Nuclear University “MEPhI”

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Who can take this course?

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