The 0 step depletion and transport keff result are different

User Support
1

Hi,
I have a problem with the depletion. Why are the result of keff between the 0 step depletion and the direct transport are so different? The following figure shows the result. Please see the result of the gas cooled fast reactor. But this unacceptable difference did not occur when I calculated the fuel pin cell. I have no idea why the keff result are so different .
I use the version 0.13.2 and ‘chain_casl_sfr.xml’.

result
result1026×559 14.5 KB

Thank you
Xu

2

Hello. I agree this is an unexpected result. The core model is for an HTGR, presumably of the prismatic type. I assume the ‘fuel pin’ is a graphite, TRISO-fueled HTGR compact surrounded by graphite.

How are you representing the regions subject to burnup in the pin model, and how are you doing this in the core model?

Excluding a bug in the code, then if the direct and 0-burnup core calculations differ, it has to be because the material compositions in effect during the transport calculations also differ, or that the boundary conditions of the system are different.

OpenMC does not do burnup - the Python API handles that. Is there a possibility that the material regions subject to burnup are being smeared prior to the 0-burnup transport step? That would explain your observations.

3

Hello. Firstly thanks for you getting back to me. But the core model isn’t for an HTGR. The core model is for a sCO2 cooled fast reactor and use the traditional rod type fuel.

I agree the result may be caused by material composition. I guss the reason leading to the unexpected result is the fuel material compositions. The fuel material for the sCO2 cooled fast reactor is UO2-10%BeO which BeO and UO2 are homogenization. The result is reasonable when I removed the BeO.( The keff result of the transport is 1.05871, the keff result of the 0-burnup is 1.05868. ) And in the previous calculation of the common fuel material (UO2/UC), the unacceptable difference didn’t appear.

But now I don’t kown how to get the correct depletion result of the UO2-10%BeO fuel. I would appreciate it if you could give me some suggestions about it.

4

Please allow me to ask a few clarification questions, to be certain I understand the issue:

  1. Please confirm: when comparing k-inf for an infinite lattice fuel pin model with /just/ UO2, you get agreement between direct transport and 0-burnup step transport. Is this correct?

  2. When re-introducing the BeO in the fuel composition, a significant discrepancy is introduced when you compare k-inf obtained using direct transport and 0-burnup step transport. Is this correct?

  3. How are you implementing the mixed composition? Is it a single material declaration with multiple isotopes added? Or are you creating two separate material instances and using Python API features to mix them?

  4. What is the fuel pin segmentation for the problem? Do you get disagreements in k-inf if there is no segmentation? That is to say to use the entire fuel pin as the burnup region?