"""
This program :
    1. Read a MCNPX output and calculate its variables.

    2. Write 2 text files, containing:
        a. Tallies, with the following outputs:
            - Conversion ratio*, **
            - Fission rate of U-235
            - Absorption rate of Th-232
            - Neutron spectrum (energy vs flux)
                *the conversion ratio is an approximation
                **uncompleted
        b. keff and its standard deviation
        
    3. Plot multiples graphs:
        a. keff
        b. conversion ratio
        c. flux (tally)
        d. cross section

@author: phitonthel 
thephitonthel@gmail.com
"""