Hello everyone !
I would also like to ask about flux distribution calculations in OpenMC.
For reactor components such as the moderator/reflector and the irradiation sites, what is the recommended way to obtain and visualize the neutron flux distribution:
Should these distributions typically be obtained using mesh tallies, regular cell tallies, or another approach in OpenMC?
Hi @DinaEl
I think that the most convenient way it is to use mesh tallies, and you could define and plot both radial and axial ones. Besides that, from 2D distribution you can take a profile along any line you want.
Hello
In OpenMC I am using mesh tallies to obtain flux distributions in a reactor model
I would like to know the recommended approach for:
Should I use a Cartesian mesh and post-process to radial coordinates, or is cylindrical meshing preferred for better accuracy and statistics
Thanks
The choice between cartesian and cylindrical coordintes is dictated by the symmetry of your problem (rectangular systems or radially symmetric ones), so the accuracy here will be higher if your choice will follow your symmetry problem.
Technically speaking, you could use cartesian coordinates with sufficient mesh refinement for any geometry, since they are more easier to process and renormalize, but you could lost on accuracy and time calculation cost…if used for strictly cylindrical geometry, where cylindrical coordinates are more suitable.
Hello and thank you for your explanation.
I want to obtain neutron flux distributions in a reactor, both in the radial and axial directions, and to evaluate the flux in different regions of the reactor (fuel, moderator, reflector, etc.).
In this case, would a cylindrical mesh tally be preferred over a Cartesian mesh for obtaining radial and axial flux distributions?
If I use a sufficiently refined Cartesian mesh, should I expect significant differences in the calculated flux compared with a cylindrical mesh tally, or mainly a difference in computational cost?
Would you recommend using a cylindrical mesh tally for this purpose, or is it better to use cell-based tallies for each region?
What is the usual approach for obtaining radial and axial flux distributions in each reactor region
which reactor geometry are you using? rectangular Fuel Assemblies or Hexagonal?
the fuel assembly geometry is hexagonal