using tallied fission distribution as the initial source particle distribution

User Support
1

it is possible to specify the initial particle source spatial distribution, throughout the domain, from the fission site distribution from an earlier simulation? and what would the syntax be?

2

Hi,

Yes, it is possible to use a source distribution from a previous simulation as the starting distribution for a new one. If you are using the Python API, it would look something like the following. When you are first generating your model, in your settings object, you could have:

settings = openmc.Settings()
settings.sourcepoint = {‘separate’: True}
…
settings.export_to_xml()

which tells it to write the source distribution separate from the normal statepoint file (that contains tally results and other metadata). Then, for a second simulation, you would tell OpenMC to use the source file that was written out previously with (assuming the previous simulation ran for 50 batches):

src = openmc.Source(filename=‘source.50.h5’)
settings.source = [src]

If you are doing this directly from XML, the above Python code would map to the following XML:

true

Best regards,
Paul

3

great, thank you very much.

4

Should I set photon_transport to true when using settings.sourcepoint = {‘separate’: True}?