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Description:

• This course covers in detail the underlying theory of nuclear fission reactors. The relevant concepts from nuclear physics, such as nuclear reactions, cross sections, nuclear data sets, and fission chain reactions, are outlined at the beginning. The general form of neutron transport equation is derived and standard approximations to this equation are discussed. Numerical methods used in solving the neutron transport equation are outlined. The one-speed one-group and multigroup neutron diffusion equations are derived. In addition to elementary analytical solutions, several numerical methods for solving the neutron diffusion equation will be presented. Neutron thermalization is discussed in detail. Reactor kinetics and control are also discussed. In addition, lecture will include more engineering subjects such as inhomogeneous reactors, cell calculations, reactivity control and fuel depletion.

Prerequisites:

• 101, 150; Engineering 117 recommended.

Textbook:

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

Computer Usage:

• About 20% of homework assignments using UNIX workstations or personal computers and FORTRAN or C computer languages. Typically an elementary finite-difference solution of a diffusion equation, and a solution of an ODE system or radioactive decay schemes and/or a point reactor kinetics problem. Students will use several modern reactor analysis computer programs (GTRAN2, CPM-2, MCNP). Course Accounts: All student will get the course computer accounts in the Advanced Nuclear Engineering Computing Laboratory (1106 Etcheverry Hall)