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Self-Shielding
REACTION RATES - h1019v1_122

Nuclear Physics and Reactor Theory Volume 1 of 2
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DOE-HDBK-1019/1-93 Reactor Theory (Neutron Characteristics) NUCLEAR CROSS SECTIONS AND NEUTRON FLUX Rev. 0 Page 17 NP-02 Nuclear Cross Section and Neutron Flux Summary Atom density (N) is the number of atoms of a given type per unit volume of material. Microscopic cross section ()   ) is the probability of a given reaction occurring between a neutron and a nucleus.   Microscopic cross sections are measured in units of barns, where 1 barn = 10-24 cm .2 Macroscopic cross section (*   ) is the probability of a given reaction occurring per unit length of travel of the neutron.  The units for macroscopic cross section are cm  . -1 The mean free path (    ) is the average distance that a neutron travels in a material between interactions.   Neutron flux (1   ) is the total path length traveled by all neutrons in one cubic centimeter of material during one second. The macroscopic cross section for a material can be calculated using the equation below. *   =  N ) The absorption cross section for a material usually has three distinct regions.  At low neutron energies (<1 eV) the cross section is inversely proportional to the neutron velocity.   Resonance absorption occurs when the sum of the kinetic energy of the neutron and its binding energy is equal to an allowed nuclear energy level of the nucleus. Resonance peaks exist at intermediate energy levels.  For higher neutron energies, the absorption cross section steadily decreases as the neutron energy increases. The mean free path equals 1/*   . The macroscopic cross section for a mixture of materials can be calculated using the equation below. *   =  N  )      +  N  )      +  N  )      +  ....... N ) 1 1 2 2 3 3 n n Self-shielding is where the local neutron flux is depressed within a material due to neutron absorption near the surface of the material.







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