By solving the transport equation in multi-group form, then you can get the solution. or using MCNP-like code and tally the flux at defined component as you wish

Dear Youqi

Are you able to suggest me documents to help me to create adequate MCNP file for flux calculations 

Best Regards

Take a look at this techreport from LANL, it should help you out:

http://www.mcnp.ir/admin/imgs/1354173299.Criticality%20Calculations%20with%20MCNP5%20A%20Primer%202nd%20Edition.pdf

Thank you

Dear Amrani,

Please see the cell calculation theme in the neutronic calculation area.

In case of "sodium fast reactors" the whole neutron spectrum is part of the "fast spectrum". Nevertheless, the energy grid depends on the use of the results. Example: the neutron irradiation damage exposition of a certain steel can be treated as the fluence of the neutrons above 0.5 MeV, while for another steel the limit may be 1 MeV, and for the third kind the dpa (displacement per atom). You should first define the purpose of the calculational results, thereafter set up the energy grid.