Critical Boron Concentration in Burnup Calculation

User Support
1

Hello All,

I want to know how to calculate the boron concentration (in ppm) change as a function of burnup in openmc?

Thanks.

2

Right now it wouldn’t be possible to do this during depletion itself. What you’d have to do is to run a critical search on the boron concentration at each depletion time point as a post-processing step. You can recreate a model with the correct material compositions by using openmc.deplete.ResultsList.export_to_materials (new in version 0.12.1) and then use openmc.search_for_keff to find the critical boron concentration. However, note that this is not the same as doing a critical boron search during each depletion time point (which itself would affect the reaction rates and evolution of the material compositions).

3

Thanks @paulromano

Instead of running depletion for the 2-month cycle at a time, I will run it for the first 20 days (let’s say) to see my system is critical or not, if no I will search for boron concentration to make my system again critical, update my materials.xml file and jump to the next burnup step. The whole process will keep continuing.

If I follow this process, I think I can address

issue, right @paulromano?

The second problem is to find out the sub-critical state and when it will happen? My plan is to run it for 2 month cycle first, then I will get a rough idea about the sub-critical state and take it into consideration in my second attempt while adjusting boron concentration to make it critical.
Is there any better way to handle this issue here @paulromano?

4

Yes, what you’ve proposed should work – basically running a separate depletion calculation for each timestep whereby you’ve already done the boron search to ensure criticality. As to your second question, because you’re controlling reactivity with boron, the boron concentration will slowly go down as the fuel is depleted. Once the fuel is sufficiently depleted, you won’t be able to reach criticality even with no boron in the system.

5

Thanks, @paulromano

I have encountered an error message

WARNING: Could not find the cell containing particle 442

I have also run it in geometry debugging mode with the openmc -g but It didn’t print any useful info.

Here’s my input file

6

@ayaz It looks like you haven’t specified a source distribution at all. By default, OpenMC will source particles at the origin, (0, 0, 0). The z-dimension in your geometry starts at z=0, so particles would be sourced directly on the boundary, which can cause issues like this. You should define a starting source that covers the entire length of your fuel bundle. See the user’s guide section on external source distributions for further information.