![]() The fission neutrons (fast-flux) in the thermal reactor are immediately slowed down to the thermal energies via neutron moderation. Distribution of kinetic energies of neutrons in the thermal reactor and the fast neutrons reactor. If we want to know the reactor core’s reaction rate or thermal power, it is necessary to know how many neutrons are traveling through the material. ![]() In other words, we do not know anything about the power level of the reactor core. These parameters influence the criticality of the reactor core. It was determined the mean free path of neutrons in the material under specific conditions. The section on the neutron cross-section determined the probability of a neutron undergoing a specific neutron-nuclear reaction. It is a scalar quantity, and it can be calculated as the neutron density (n) multiplied by neutron velocity (v). The neutron flux density, Ф, is the number of neutrons crossing through some arbitrary cross-sectional unit area in all directions per unit time.
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