Finding absorption and fission cross-section from openmc for a simple 1d geometry

So i have a simple 1d geometry, which i have solved analytically, i want to solve the same thing in openmc, and see what holds true.

However, for the analytical solutions i have assumed constants for v, k, E_a and E_f so i was wondering if there was a way in which i could find those constants from openmc

I am assuming you analytically solved the neutron transport equation for one energy group. If this is the case, you can check your results by running OpenMC in multigroup mode with just one energy group. In order to do this you may want to generate single group macroscopic cross sections through OpenMC or use your own nuclear data.

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