Thank you for reading my question.

I want to calculate "energy averaged" one group cross-section by using MCNP for substituting ORIGEN-2 LIB(just one energy group cross-section).

And I would use σ=R/(N*Φ).

σ is the averaged cross section (barn) (This is what I want to know)

R is the reaction rate (/cm^3) (maybe calculated by F14 tally and FM14)

N is the target atom density (maybe atom/barn-cm)

Φ is energy-integrated neutron flux (maybe calculated by F4 tally)

So, if I want to calculate 59Co (n,r) 60Co averaged cross-section, what FM is correct, FM4 (1.0 27059 102), FM4 (c 27059 102) or etc? ; c is the atomic density(atom/barn-cm) of Co-59 of M1.

Under text is simplified data card of MCNP input text of my research.

I want to know the energy averaged cross-section of M1's Co-59 activation reaction.

C M1 is SUS-304

M1 24000.50c -0.19

25055.66c -0.02

26000.55c -0.694

28000.50c -0.095

27059.66c -0.001

C Material for Tallying

M27059 27059.66c -1

F4:n 1

FC4 Neutron Flux of cell 1 Calc. F14:n 1

FC14 Co-59 Activation Reaction Rate of cell 1 Calc.

FM14 (1.0 27059 102) or (c 27059 102) or (1.0 1 102) or (c 1 102) or etc...

It may be cumbersome, but it would be very helpful if you answered.