Neutron Flux

  • Thread starter NukeLion
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when we calculate the neutron flux in finite medium using sn method for steady state neutron transport equation, it gives us some numbers up to 1. I am sure its not the real flux, can someone explain how we can calculte the real flux using sn method.
 

Astronuc

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when we calculate the neutron flux in finite medium using sn method for steady state neutron transport equation, it gives us some numbers up to 1. I am sure its not the real flux, can someone explain how we can calculte the real flux using sn method.
If values are going from 0 to 1 for a value, then it is probably normalized or represents a probability. Then the question is to what quantity I the value normalized. Perhaps the reaction rate or neutron production rate/density, which relates to the fission density.

One can solve for a neutron distribution, but the actual value depends on fission density. A finite critical fission system can have any power level up to some limit, e.g., 1 W to 1 GW. Criticality means steady-state. The local fission density will depend on the relative fission cross-section in comparison to other absorption cross-sections.
 
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Thanks Astronuc for your reply. So is it means that we need to multiply with power to get the real flux?
 

Astronuc

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Thanks Astronuc for your reply. So is it means that we need to multiply with power to get the real flux?
Correct, there would have to be some local or average (integrated) power from which to obtain a real flux. In conjunction with power and temperature (assuming some heat transfer), the temperature would have to be consistent with the resonance (doppler) broadening and moderator density.
 

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