MCNP to OpenMC SDEF conversion tool

With this tool, you can convert a fusion neutron source from MCNP to OpenMC.

Currently only cylindrical SDEF with radial and vertical dependent distributions is supported

• At the moment, work has been tested only with fusion neutron sources based on cylindrical geometry with radial and vertical dependent distributions.
• An example of such a source for MCNP can be found in the /notebooks/data folder.

There isn't any methodical V&V on this tool. Use at your own risk.

Installation

pip install openmc_sdef_parser*.whl

Usage

• Import package:

  import openmc_sdef_parser as parse

• Link variable with the path to SDEF file:

  sdef = parse.make_openmc_source('path/to/sdef')

• Create settings object:

  settings = openmc.Settings()

• Add source to settings attribute:

  settings.source = sdef

Attributes

• sdef = parse.make_openmc_source('path/to/sdef')

  • sdef is an openmc.IndependentSource object

• sdef_file is the path to the SDEF file

• normalize is whether to normalize the source per 1. Default is True

• exclude_first is whether to exclude the first z bin (vertical distribution). It may be useful in some cases. Default is False, i.e. include the first z bin and exclude the last z bin.

Plotting

• If you want to plot your SDEF:

  sdef = parse.make_openmc_source('path/to/sdef')
parse.plot_source(sdef)

• Options:

  • By default, the source is normalized to 1.
    If you want to plot the intensity from min to max, set intensity_norm = False
  • You can change the color map with cmap and the interpolation method with interpolation

  parse.plot_source(sdef, intensity_norm = False, cmap = 'viridis', interpolation = 'bicubic')