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I am trying to obtain the integral neutron flux based on the results obtained with MCNP (neutron spectrum calculation) for each energy bin (51 neutron energy bins). I have seen in many papers the calculation of the differential neutron flux multiplying the neutron flux results of MCNP by each energy bin; however, I can not figure how to obtain the integral neutron flux. Unfolding codes like SANDP or STAY´SL report integral neutron flux for each energy bin. Hope you can help me with my doubt. Thank you very much.

Best regards

Juan Galicia-Aragon