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Nuclear Reactors, BAU, 1st Semester, 2007-2008 (Saed Dababneh). 1 Review Test Consider thermal neutrons in natural uranium (19.04 g.cm -3 ), a)What is the fission, capture and total absorption macroscopic cross sections? (0.2026, 0.1652, 0.3678) cm -1 b)What is the mean free path for those neutrons? (2.72 cm) c)What is the average distance the neutron travels before it causes fission (if it does)? (4.94 cm) d)What is the neutron flux and neutron density required to produce 1 W of power per cm 3 ? (1.53x10 11 cm -2 s -1, 6.955x10 5 cm -3 ) e)Compare this neutron density to the atom density of the fuel. (4.8183x10 22 cm -3 ) Conclusion: Probability for n-n reaction << than that for absorption in the fuel. This makes calculations EASIER…!!!

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Nuclear Reactors, BAU, 1st Semester, 2007-2008 (Saed Dababneh). 2 Review Test Why do you think such a reactor that uses natural uranium as a fuel (like CANDU) is larger in volume than a PWR that uses enriched uranium? Why should we worry about neutrons that leak out? shield attenuate Boron is a common shield against thermal neutrons. Estimate the thickness of boron required to attenuate a neutron beam to 0.1% its intensity. Compare to lead.

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Nuclear Reactors, BAU, 1st Semester, 2007-2008 (Saed Dababneh). 3 Review Test Describe thoroughly the s -wave scattering of neutrons from 1 H. Emphasize on the energy distribution and the angular distribution in the CM as well as in the lab systems. Calculate the mean number of fission neutrons produced per thermal neutron if the fuel used was enriched uranium (5%). Comment on the effect of your result on the choice of the moderator.

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Nuclear Reactors, BAU, 1st Semester, 2007-2008 (Saed Dababneh). 1 Multi-group Model Wide neutron spectrum. One-group, two-group? Should be generalized.

Nuclear Reactors, BAU, 1st Semester, 2007-2008 (Saed Dababneh). 1 Multi-group Model Wide neutron spectrum. One-group, two-group? Should be generalized.

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