How openmc deal with cross-sectional data

User Support
1

Hello!

   The upper limit of many elements' cross-sectional data in the

ENDF_B-VII.0 is 20MeV. If some of the neutrons’ energy is exceeds
20MeV, how does OpenMC handle it during calculation? Is it directly
killing this particle? Where can I find the relevant instructions in
the OpenMC manual?

Thanks

2

Hello,

As you can see from a typical output from running openmc, you’ll get a message like this:

 Maximum neutron transport energy: 20000000 eV for N15

If a particle is sampled above the upper bound energy, a new energy is retried. So, the particles are not killed, it’s instead that the energy distribution will be truncated and re scaled to fall in the allowable range. You can see the relevant behavior in src/source.cpp. Do you think this should be specifically noted in the documentation?

3

Thanks for taking the time to answer my question.Your help is greatly appreciated.