MCNP output file interpretation

[kir]

I am running (a very basic) simulation of the proposed LIFE concept reactor at LLNL as part of my MSc thesis. What I hope to achieve is to calculate the fission energy gain from a fissile blanket surrounding a source of fusion neutrons (ie D-T pellet blasted by lasers)

The problem summary table in the output lists the neutron creation and neutron losses due to various interactions. My question is whether it is correct to take the 'loss to fission' as an indicator of how many fissions occur in the system, and if so would the loss to fission weight per source particle be the correct value to use in calculating the total fission energy gain?

 

[QuantumPion]

I'm not sure what that loss to fission parameter is. Check your MCNP manual under the section for tallys. I believe there is a specific tally for power or power density (f or fm or something).

 

[Devesh Raj]

Dear Kir
If you solved your problem please enlighten me also.
Best wishes

 

[MCNPX]

"Loss to fission weight" gives the number of fissions, however one should be careful with normalization. Also, be sure that you use the correct value of average energy per fission for your system - it can be tricky in some cases.

I am running (a very basic) simulation of the proposed LIFE concept reactor at LLNL as part of my MSc thesis. What I hope to achieve is to calculate the fission energy gain from a fissile blanket surrounding a source of fusion neutrons (ie D-T pellet blasted by lasers)

The problem summary table in the output lists the neutron creation and neutron losses due to various interactions. My question is whether it is correct to take the 'loss to fission' as an indicator of how many fissions occur in the system, and if so would the loss to fission weight per source particle be the correct value to use in calculating the total fission energy gain?