Transcript Document

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?
b) What is the mean free path for those neutrons?
c) What is the average distance the neutron travels
before it causes fission (if it does)?
d) What is the neutron flux and neutron density required
to produce 1 W of power per cm3?
e) Compare this neutron density to the atom density of
the fuel.
Conclusion: Probability for n-n reaction << than that
for absorption in the fuel. This makes calculations
Nuclear Reactors, BAU, 1st Semester, 2007-2008 (Saed
1
EASIER…!!!
Dababneh).
(0.2026, 0.1652, 0.3678) cm-1
(2.72 cm)
(4.94 cm)
(1.53x1011 cm-2s-1, 6.955x105 cm-3)
(4.8183x1022 cm-3)
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?
• 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.
Nuclear Reactors, BAU, 1st Semester, 2007-2008 (Saed
Dababneh).
2
Review Test
• Describe thoroughly the s-wave scattering of neutrons
from 1H. 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.
Nuclear Reactors, BAU, 1st Semester, 2007-2008 (Saed
Dababneh).
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