Thanks for this awesome simulation tool.

As a test I filled a reflective cube (all 6 sides reflective) with metallic density natural Uranium (235-U : 0.007172 the rest is 238-U … No 16-O…) and calculated keff for infinite geometry.

Results are consistently keff=0.425

However I read in ADS literature that MCNPX 2.7 result is 0.816 https://www.sciencedirect.com/science/article/abs/pii/S0360319915002335

Other literature suggests around 0.6 k-inf value.

The difference is significant, and I haven’t had the time and patience to review PFNS, (n, xn), inelastic, etc… effects (I remember that I had difficulties estimating inelastic energy loss when I was implementing a simple MC code before openmc was written and became way superior).

Density does not effect keff at infinite geometry, but I tried different (eg. 12.0 … 18.9 ) with same keff=0.425 result. Traces of 234-U present or not makes no significant difference.

Which is the most suitable data library for 238-U with “very fast” (few steps from PFNS) spectrum? Currently using ENDF/B-VII.1

I have no way to verify with physical experiment.

Which is the most relevant (finite geometry) assembly that is, or could be used for validation of 238-U fast-spectrum behavior?

What k_inf is expected for natural Uranium?

Thanks a lot,

Marcell

Hi Coupled, welcome to the community

I am making a simple case for all three materials (natural Th, natural U, and depleted U) using material composition from PNNL Compendium of Material Composition Data for Radiation Transport Modeling | Report | PNNL

natural.ipynb (101.1 KB)

and I got the same kinf as you when using ENDF/B-VII.1

So I think the kinf calculated by Sahin that you mentioned is using a different approach since that manuscript talked about ADS with its specific geometry and 1GeV proton. I think that the study uses fixed source calculation instead of eigenvalue calculation.

regarding your question

Which is the most suitable data library for 238-U with “very fast” (few steps from PFNS) spectrum? Currently using ENDF/B-VII.1

I think you can use ENDF/B-VII.1 at first, and then compare it with other nuclear data libraries: VIII.0, JEFF, etc. I think all these nuclear data cover incident neutron data for up to 20MeV, some photon data and also a wide range of temperatures, you could check it on the official openmc nuclear data library

https://openmc.org/official-data-libraries/

about this question

Which is the most relevant (finite geometry) assembly that is, or could be used for validation of 238-U fast-spectrum behavior?

I have no experience in this case, but I have done the criticality primer from MCNP on openmc, and one of their cases (Example 2.5) uses natural uranium as a reflector for their plutonium cylinder if you want to check it. I think the neutron spectrum for that example case is within the fast neutron range.

Each case has a reference keff solution to make sure that we have a good model.

sorry