Is MCNP capable of scoring fission spectrum?

[froztiz]

Hi everybody,
I am trying to score fission spectrum in MCNP for a kcode calculation. I would like to check at which energy neutrons produced by fission are generated. I have no idea how to perform since tallies are usually volume or material dependent and I just want to build a spectrum containing only the neutrons created by fission.
Any idea on how to proceed?

 

[TheNE]

I cannot advise on how to tally the energy right upon neutron "birth", but you can tag the neutron tally to only be scored for specific reactions in specific cells/materials. This material begins in the MCNP 6.2 User Manual at 3-245 or 302/793 on the PDF copy.

Edit: The tag will act as a "barcode" upon the history that is a result of the specified reaction in the specified cell/material. Your tally will show the contribution from that specific tag and any other specified tags. This is quite useful when you're wanting to examine contributions from specific nuclides to a tally.

 

[Astronuc]

Hi everybody,
I am trying to score fission spectrum in MCNP for a kcode calculation. I would like to check at which energy neutrons produced by fission are generated. I have no idea how to perform since tallies are usually volume or material dependent and I just want to build a spectrum containing only the neutrons created by fission.
Any idea on how to proceed?

As far as I know, the fission neutron energy spectrum is an input that one takes from a nuclear data library: ENDF, JEFF or JENDL.

See - https://www.osti.gov/servlets/purl/1465749

Accurate nuclear data are important for reducing the uncer-2tainty of nuclear simulation codes such as Monte Carlo N Par-3ticle (MCNP)[1] and GEANT[2]. These codes require cross-4sections and other nuclear data such as energy spectrum for5prompt fission neutron emission in order to simulate nuclear6processes.

See also - https://uknowledge.uky.edu/cgi/viewcontent.cgi?article=1065&context=physastron_etds