for the neutron activation ,  i think you should calculate reaction rate 

Thanks for your response but I haven't FL card, how can I writ it's code?

Sorry I mean not  FL but FM card. So is the previous  command  available with mcnp?

FM card is an entry to MCNP code in input file.

For exemple (n,gamma) reaction rate in a foil that has been defined as cell number 1 and contains material number 10 is calculated by introducing the following two lines in data block of MCNP input:

F4:n 1

FM4 C 10 102 

where C refers to atom density of the target (see at MCNP manual for the signe of C)

One way is to calculate a detailed neutron energy spectrum phi(E) with MCNP and then to pass the spectrum to the FISPACT-II activation code.

Alternatively, use MCNP6.2 and BURN the material. This will give you the {Nuclide, Activity} inventory matrix for the activate material. Next, use the code MCNP_SOURCE_APP to convert this into a spectroscopially accurate SOURCE TERM for MCNP.

BURN card is very suitable to activate material with MCNP. But it can be used only in kcode calculation.