MG cross section error

User Support
1

Hallo everyone,

I have created a MG cross section data library as mgxs.h5.
when I try to run depletion using this I get the error below, eventhough I have defined a path to my mgxs data like this:
os.environ[‘OPENMC_MG_CROSS_SECTIONS’] = ‘/dss/dsshome1/0C/ge25kij2/openmc_13_3/openmc/mgmc/dep_test/mgxs/mgxs.h5’

can anyone help me with this problem?

RuntimeError: No mgxs.h5 file was specified in materials.xml or in the OPENMC_MG_CROSS_SECTIONS environment variable. OpenMC needs such a file to identify where to find MG cross section libraries. Please consult the user’s guide at https://docs.openmc.org for information on how to set up MG cross section libraries. Abort(-1) on node 0 (rank 0 in comm 0): application called MPI_Abort(MPI_COMM_WORLD, -1) - process 0

2

I think OpenMC cannot perform burnup calculations in multi group mode. You can calculate the transport of each burnup step by using continuous energy to calculate the nuclear density for each burnup step