FLUKA: Help

[foucher <yann.foucher@psi.ch>]

Dear FLUKA users,

I'm working on a solid Lead-bismuth target surrounded by Light water.
Inside the water are located some plexiglass tubes and gold foils in
order to detemine
the neutron thermal flux.
I succeed to get the thermal neutron flux and now I would like to get
the reaction rate:
The neutron flux multiplied by the neutron capture cross section
Au-197(n, gamma) Au-198.
My problem is the following:
I have found the neutron capture cross section under a 640 multigroup
form (from 10^-14 GeV to
0.0196 GeV).
The neutron flux given by FLUKA contains 72 groups from 10^-14 GeV to
0.0196 GeV.
In order to multiply the cross section with the neutron flux, I would
like to know if
there is a definition of the Au-197(n, gamma) Au-198 cross section with
the same energy bin structure
(i.e: 72 groups) with the same limits of bins.
Thanks in advance

Yann FOUCHER
Spallation Neutron Source
Paul Scherrer Institut