Flux distribution and unit of reaction rate for the research reactor fuel assembly

[umarov]

Dear Users and Experts,

I have designed an IRT-4M fuel assembly for a research reactor and am seeking assistance in plotting 2D and 3D neutron flux distributions as in example ( Jupyter Notebook Viewer (nbviewer.org)

). Despite multiple attempts, I have been unsuccessful. Could someone provide an example on my code to help me learn how to obtain and visualize flux distributions?

Additionally, I have plotted a fission reaction rate distribution but am uncertain about the units. I expected the reaction rate to be on the order of 10^4 10^6. How can I convert the reaction rate to real values?

Thank you for your assistance.

IRT-4M.ipynb (136.8 KB)

[wahidluthfi]

I think you can read other discussions regarding tally normalization, Paul has been describing tally normalization properly in this discussion

NOrmalizing Tally to get Flux value

The tally is not normalized by volume, so the units are something like [neutrons-cm/source]. This can be understood quite easily: in a collision estimator tally, OpenMC accumulates 1/Sigma_t which has units of cm, and similarly for a tracklength tally, OpenMC accumulates the tracklength in cm. When you divide by volume, you get [neutrons/cm^2-source]. The normalization factor Pnu/(Qk) gives you [J/secneutrons/fission/(J/fissionneutrons/source)] = [source/sec] so that when you multiply the flux b…