NE 250 - Nuclear Reactor Theory

Objectives:

1.    To understand fundamentals of nuclear reactor physics;
2.    To understand neutron transport theory, neutron diffusion theory, and their relationship; 
3.    To understand neutron behavior in nuclear reactors;
4.    To develop physical and computational skills to analyze and solve nuclear engineering problems.

Description:

This course illustrates the underlying theory of nuclear fission reactors and discusses related nuclear engineering subjects.  Commonly used nuclear reactor systems are introduced.  The fundamental concepts in nuclear reactor physics such as nuclear cross sections, nuclear reactions, and criticality of nuclear fission chain reactions are reviewed.  The neutron transport equation is derived and approximations to this equation are discussed.  Numerical methods used in solving the neutron transport equation are discussed.  The relationship between neutron transport theory and neutron diffusion theory are discussed.  The analytic and numerical methods to solve the neutron diffusion equation are discussed.  Neutron slowing down, lattice physics and core physics methods, adjoint equation and perturbation theory, and reactor kinetics are discussed.  Fuel depletion, radiological source, and decay heat are also discussed.

Text: 
J. J. Duderstadt & L. J. Hamilton, "Nuclear Reactor Analysis," Wiley (1976)

References:
1. A. Henry, "Nuclear-Reactor Analysis," MIT Press (1976)
2. Bell and Glasstone, “Nuclear Reactor Theory,” Van Nostrand Reinhold (1970).
3. J.R. Lamarsh, "Introduction to Nuclear Reactor Theory," Addison-Wesley (1966)
4. K.O. Ott and W.A. Bezella, "Introductory Nuclear Reactor Statics," American Nuclear Society, LaGrange Park, IL (1983).
5. W. Stacey, "Nuclear Reactor Physics", Wiley (2001)


Prerequisites: NE 101, NE 150; (E 117 recommended) or consent of instructor.

  Course Computer Accounts and Computer Laboratories:
Course Outline
General description of nuclear fission reactor systems and current status and trend in nuclear power plant development
Review of nuclear force, nuclear structure, nuclear stability, nuclear radiations, nuclear cross sections, evaluated nuclear data libraries, resonances, Doppler broadening, neutron transport, neutron diffusion, neutron flux, neutron current, nuclear reactions, energy release per fission, multiplication factor and criticality, delayed neutrons, nuclear fuel, conversion, breeding, interaction of radiations with matter, radiation damage, and four- and six-factor formulas for neutron multiplication
Derivation of neutron transport equation, integral transport equation, PN equations, P1 approximation, and analytic solution to source-free infinite medium one-speed transport equation
Discrete ordinates SN method, collision probability method, method of characteristics, and Monte Carlo method
Diffusion approximation of neutron transport equation, derivation of Fick's law with transport correction and neutron diffusion equation, and multi-group neutron diffusion theory
Classical nodal method and modern nodal methods (analytic method, Green’s function nodal method, and nodal expansion method)
Elastic scattering mechanics, energy loss, average logarithmic energy decrement, effect of inelastic scattering, collision and slowing down densities, and resonance absorption evaluation
Fuel cell fine-group thermal spectrum calculation, fast and resonance fine-group spectrum calculations, energy condensation for few-group cross section generation, fuel lattice few-group spectrum calculation, lattice-homogenized cross section generation, and effective one group neutron diffusion theory method for core simulation
Derivation of adjoint equation, physical meaning of adjoint flux, and estimate of reactivity change using first-order perturbation theory
Space-time reactor kinetics, λ-mode reactor kinetics, ω-mode reactor kinetics, point reactor kinetics, xenon transient, reactor operation transient such as control rod move, core flow change, and pump trip
Fuel depletion evaluation, exposure-dependent fission products, actinides, and activation products activity and decay heat evaluation


Back to NE 250 Home page