Given that the average number of fast neutrons emitted followingthe thermal-neutron induced fission of 235U is 2.42 per fissionevent; use the following data to calculate the mean number offission neutrons produced per initial thermal neutron in a largevolume sample of
(a) pure 235U (b) natural uranium 238U, and (c) uranium enriched to3% in the 235U isotope.
Note: The microscopic absorption cross section for 235U is 694barns. The cross section for 238U is 2.71 barns. The fission crosssection for 235U is 582 barns. Natural uranium contains 0.7%235U.
Comment on you results in terms of operation of a thermal reactorof finite size.
