OpenMC to Model Point Source Problem

User Support
1

Hi,

Is there a way I can use OpenMC to do a neutron sheilding problem? I want to have an isotropic point source of neutrons, a paraffin shield, and a detector on the other end of the parafffin.

I want to see how the neutron spectrum changes as I change the width of the paraffin shield.

2

Hi Terry,

See the section on the manual regarding Fixed Source problems: http://mit-crpg.github.io/openmc/usersguide/input.html#fixed-source-element
You will want to use that vice the element.

3

Ah, I see. thank you!