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Slowing Down of Neutrons
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absorption cross section absorption in slowing albedo approximation atoms average number beta decay boundary conditions calculate control rods critical equation cubic centimeter delayed neutrons distance elastic collision energy E0 equa equation 9-4 excess reactivity expression fast neutrons fission neutrons fissionable material function given grad nv graphite half-thickness infinite integral leakage linear mean free path medium moderator multiplication constant neutron density neutron diffusion neutron energy neutrons are absorbed neutrons are produced neutrons emitted neutrons produced neutrons slow nucleus number of fission number of neutrons nvth Oak Ridge obtain one-group theory original energy pile and reflector pile equations probability problem prompt neutrons q neutrons Q(Eth quantity ratio reactor reflector savings reflector thickness scattering cross section second per cubic sinh slowing-down density solution of equation solve spherical thermal neutron absorbed thermal neutron flux thermal pile tion transport mean free V-Am velocity