Recent content by mhovi

  1. M

    MCNP Flux and Power Calculation

    During a reactor assembly calculation, I need to determine axial and radial flux distribution over the surface. When I use F2 and F4 tally I get some value with unit 1/cm**2 What does the value means, neutron flux is supposed to be in the 10^14 range but output values are 10^2 range. Can anyone...
  2. M

    MCNP6.2 BURN Problem Uranium Dioxide 4.2% Enrichment

    uranium dioxide with 4.2% enrichment c Cell Cards 101 2 -0.0003922 -7 -5 6 imp:n=1 vol= 0.26195 tmp= 1.0109E-7 201 1 -10.55 7 -8 -5 6 imp:n=1 vol= 8.84672 tmp= 1.0109E-7 301 2 -0.001598 8 -9 -5 6 imp:n=1 vol= 0.288775 tmp= 1.0109E-7 401 3...
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