I believe the cause of this issue has now been identified.
During depletion calculations, OpenMC adds all nuclides in the depletion chain that have available neutron cross-section data to the transport calculation at very small concentrations. After commit e57916896934ed0530c41f06af11df66db9a4240, the upper neutron transport energy is determined as the minimum upper energy bound among all loaded nuclide cross-section datasets.
In this example, the full depletion chain contains Be7, whose neutron cross-section data extend only up to 8.1 MeV. Therefore, when the full chain is used, the maximum neutron transport energy is limited to:
Maximum neutron transport energy: 8100000 eV for Be7
This changes the transport calculation and consequently affects the calculated keff.
In contrast, when the simplified depletion chain is used, the output is:
Maximum neutron transport energy: 20000000 eV for Na23
Thus, the difference between the transport and depletion results is not caused by the very small Be7 concentration itself, but by Be7 limiting the global neutron transport energy range to 8.1 MeV.
I suggest paying close attention to the following message in the OpenMC output:
Maximum neutron transport energy: ...
It can help identify whether a nuclide introduced through the depletion chain is unintentionally restricting the transport energy range.
The following function can also be used to check which nuclides in the depletion chain have neutron cross-section data with an upper energy limit below 20 MeV:
test_max_energy.py (1.5 KB)