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summary of transport capabilities (e.g. flexibility of tallies, homogenization, continuous energy or multi-group, etc.)
High-quality MC code that can run both in CE and MG mode. Good support for tallies of various reactions, and with filters for materials / cells / nuclides, etc. No fission matrix as of the time of this writing
summary of geometry modeling
CSG
Development
Are you willing to develop this interface? If not, do you know someone with the knowledge to do it? Please link them in the issue
Yes.
Will you (or the developer linked previously) help maintain it?
Yes
How can this be tested? Writing input files can be done w/o running the executable, but may be more error prone unless we can run the solver in the test environment
Can be installed with conda or from source directly. Writing input files would surely work but with some of the same issues I've been seeing with the Serpent writer (test fails, materials aren't properly updated/reset, causing the test files to be differently only in a material id). Can also run the full sequence
The text was updated successfully, but these errors were encountered:
Overview
Please describe the solver, in particular
docs.openmc.org
https://github.com/openmc-dev/openmc
High-quality MC code that can run both in CE and MG mode. Good support for tallies of various reactions, and with filters for materials / cells / nuclides, etc. No fission matrix as of the time of this writing
CSG
Development
Yes.
Yes
Can be installed with conda or from source directly. Writing input files would surely work but with some of the same issues I've been seeing with the Serpent writer (test fails, materials aren't properly updated/reset, causing the test files to be differently only in a material id). Can also run the full sequence
The text was updated successfully, but these errors were encountered: