You need a neutron spectrum and damage cross section.

Try to obtain Neutron Flux from the simulation output. Therefore, you have to set your tally mesh considering the neutron record.

In order to have a "1-MeV Equivalent Fluence" you will need to specify the material for the damage metric. ASTM Standard E722 provides the damage equivalent metrics at 1-MeV for both silicon and gallium arsenide. As chair of the ASTM E10.07, I produced a report on how the recent changes (2014) would impact the calculation of damage metrics:

"Silicon Damage Response Function Derivation and Verification: Assessment of Impact on ASTM Standard E722"

You can find it in my profile. The Appendices contain the response functions (E/EM cards) for MCNP to convert a fluence to a 1-MeV(Si) equivalent fluence. I recommend examining the E722 standard to grasp what the function is actually doing.