Integral Neutron Flux: Getting Results with MCNP - Juan Galicia-Aragon

[Juan Aragon]

Hello everyone

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

 

[rpp]

In the nuclear engineering field, I've never seen anybody use the differential neutron flux. I'm not saying there isn't an application, I've just never seen one. Maybe there are some applications in the nuclear physics area?

The flux you calculate in MCNP is usually the integral flux (total flux in a group). I'm not sure exactly what you are trying to do, but you can usually get the integral flux by using a type 4 tally, and then use the "e" multiplier to define different energy bins.