| |
| | ne150 | Nuclear Engineering, UC, Berkeley |
 | | review those aspects of neutron interactions with matter that are pertinent to understanding the establishment of a chain-reaction and of the neutron space- and energy-distribution in the nuclear reactor core. |
 | | calculate spectrum-averaged microscopic cross-sections for thermal neutrons, macroscopic cross-sections for a single isotope and for a mixture of isotopes, reaction probabilities, mean-free-path, mean time for collision, mean energy loss per elastic collision. |
 | | Introduction to neutron diffusion theory: neutron flux and current, equation of continuity, Fick's law, transport corrections; the diffusion equation for monoenergetic neutrons, boundary conditions; elementary solutions of the steady-state diffusion equation, solutions for multiplying media, multi-group diffusion equations; solution of the two-group diffusion equation. |
| www.nuc.berkeley.edu /courses/classes/ne150.html (1014 words) |
|