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Neutron Flux

  1. Dec 24, 2014 #1
    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.
     
  2. jcsd
  3. Dec 24, 2014 #2

    Astronuc

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    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.
     
  4. Dec 31, 2014 #3
    Thanks Astronuc for your reply. So is it means that we need to multiply with power to get the real flux?
     
  5. Dec 31, 2014 #4

    Astronuc

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    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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