Dear OpenMC experts,
I have running an ADS problem on fixed source simulation with an external source file. The running came out fine, but I got only the value of fluxes. The fission rates are all zeroes, as shown in the figure below. Is there anything wrong, or what causes these fission rates to become zeroes?
Here is my fluxes plot and fission plot respectively.
ads_data.h5 (7.9 KB) geometry.xml (2.5 KB) materials.xml (3.5 KB) settings.xml (276 Bytes) tallies.xml (343 Bytes)
If I add a material filter to your flux tally for the material in question, the flux result becomes 0 indicating that no particles are actually making it to the material that you are tallying fission for.
Dear @paulromano ,
According to my source fluxes plot, the sources are distributed in a ring shape at the center, which causes no interaction with other materials in the outer range? If I generate a source file that has fluxes cover all over my geometry, will this help for solving the problem?
Yes, this would certainly help ensure that particles reach other areas of the problem, but I would question whether this is what you want. Usually in a fixed source problem, you’re trying to model a physical source and don’t have the freedom to put a source wherever you want.
Now I’m creating a source file .h5 from FLUKA (impinging the PbBi target with 60MeV proton beam and obtained the distributed neutron information). I’m not sure where should the source located, is it just on the target surface (cylindrical target)? But, the result of the obtained neutron source at the cylindical target surface is shown aforementioned, the fission rate become zero. There is no particle indicating the materials in the subcritical core region as you mentioned. Therefore, I suppose to obtain the new neutron source distributed either in the volume of the target or all over the geometry(including subcritical core). please correct me.
Thank you in advance,
thanapong
To me, it sounds like you’d want your source to be distributed in the volume of the target because that is where the neutron-producing reactions are taking place.
