Question about MCNP output table of burn card

[Dedu]

neutronics and burnup data

step duration time power keff flux ave. nu ave. q burnup source
(days) (days) (MW) (GWd/MTU) (nts/sec)
0 0.000E+00 0.000E+00 4.800E+02 1.06832 2.200E+15 2.933 209.103 0.000E+00 4.203E+19
1 5.000E+02 5.000E+02 4.800E+02 1.03163 2.328E+15 2.934 209.107 2.901E+01 4.204E+19
2 5.000E+02 1.000E+03 4.800E+02 0.99949 2.460E+15 2.935 209.109 5.803E+01 4.205E+19
3 5.000E+02 1.500E+03 4.800E+02 0.97140 2.604E+15 2.936 209.108 8.704E+01 4.207E+19
4 2.240E+02 1.724E+03 4.800E+02 0.95903 2.661E+15 2.937 209.108 1.000E+02 4.207E+19

After running the calculation associated with a burn card, MCNP shows this table in the output file. What is the meaning of the FLUX column (2.200E+15)? The manual says that it is an system averaged energy integrated flux, but I don't really get it nor can I link it to other values from this table or elsewhere.

As can be seen from the table, I have run the calculation with 4 time steps of 500, 500, 500, 224 days with constant power of 480 MW (pfrac = 1 for each time step) to get to a maximum burnup of 100 GWd/MTU. I know the relation between power, average ν, average Q and source. However, I cannot relate the flux to the other quantities.

Thank you

 

[rpp]

The flux is the neutron density times the neutron velocity. You can multiply the neutron flux by the cross sections to determine the reaction rates (find the number of absorptions, number of fissions, etc.). The number of fissions times the Q value will give you the total power.
Was there another type of relationship you were expecting?

You don't mention anything about what type of system you are solving, but your depletion step size seems large. You will need to run smaller steps to get more accurate results.

 

[Dedu]

So if i multiply the flux calculated by MCNP in this table by the macroscopic XS for fission and by the total volume of fissionable material i should get the total neutron source, that is the last column from my table?
Φ [n/cm²s] × Σ_m[1/cm] (= σn) × V [cm³] = S (Total source)

Is that correct? Using the XS value of Pu-239 (one of the heavy nuclides present in the fuel in higher amount) fission by 1 MeV neutrons (kcode source = Watt spectrum, no moderator) in the above calculation gives a value pretty close to the 4.207E+19 n/s that MCNP shows in the last column (I get 1 order of magnitude higher but maybe that is due to improper approximation value of the fission XS).

Thanks for your answer btw.

 

[rpp]

Yes, that sounds correct. You will need the correct macroscopic cross section that has been averaged over all energy and space.
But if you had the one-group cross section, it would be correct.