hi all
I made an example of using code OpenMC ,which I have used the fixed source, it is executable but keff and all other values of the display is zeros, even though I have uses the neutron source which is not acceptable? is that there is someone who is view this problem
In your settings.xml file, you need to use the <fixed_source> input element instead of the element. The user’s guide describes the format for <fixed_source> here:
http://mit-crpg.github.io/openmc/usersguide/input.html#fixed-source-element
As an example, you could have
<fixed_source batches=“100” particles=“1000” />
which would run 100 batches of 1000 particles for a total of 100,000 particles.
Best,
Paul
Hi Josh, could you clarify your question for me slightly? Do you want to run a fixed source problem or an eigenvalue problem with an initial starting source? The difference being that a fixed sourced problem means you have a starting neutron source and you want to know how it propagates through and interacts with the problem geometry. An eigenvalue problem is the classic solve for the neutron multiplication factor k_eff and the associated flux.