Depletion question

[Yue JIN]

Hi All,

I ran a simulation of pin power of a 17*17 assembly. It takes several minutes to finish a simulation with 1000 batches and 4000 particles per batch. But when I run a depletion of the same model in "energy deposition" mode, it takes about 2 hours to finish a state point.

I'm not familiar with the methodologies of depletion. Why it takes much longer time?

Then I tried casl pwr depletion chain instead of endfb71. This option halves the computation time, but still time-consuming.

Does energy deposition mode have much influence on time? I tried fission q mode and set fission-q manually ({"U235": 2.02E+02,"U238": 2.12E+02, "Pu239": 2.11E+02,"Pu240": 2.14E+02,"Pu241": 2.14E+02}). The simulation exited after calculating the first state point and threw the following error: 

No energy reported from OpenMC tallies. Do your HDF5 files have heating data?

An exception has occurred, use %tb to see the full traceback.

SystemExit: 1

How to solve this problem?

Thanks in advance,

Yue.

[Jiankai YU]

Hi Yue,

The time difference is caused by the different numbers of nuclides loaded during the initilization.

For example, in pure neutron transport, only the nuclides you added in your materials.xml will be loaded. However, in the case of depletion simulation, all nuclides in the chain file with neutron-induced reaction cross section in cross_sections.xml will be loaded. And the cross section lookup (increases with numbers of nuclides) takes significant portion in Monte Carlo particle transport. For sure, you could try the simplified chain file you mentioned with MPI or OpenMP parallel computation to improve the efficiency. 

Anyway, I prefer 100 batches and 40,000 particles per batch in this case. 

 

As for the fission-q problem, it might be caused by the mistake in unit (eV). But I am not sure unless you provide more details of the model. 

https://github.com/openmc-dev/openmc/blob/develop/openmc/deplete/operator.py 

Best,

Jiankai

[Paul Romano]

To add on to what Jiankai said -- if you are using the "energy deposition" mode for depletion, you need to make sure you are either 1) using the "official" data library from openmc.org or 2) generated a library yourself with openmc.data.IncidentNeutron.from_njoy. When libraries are generated for OpenMC, we put special heating data in the HDF5 that won't exist if you just convert existing ACE files. If this special heating data does not exist in your data library, it will result in the error message you are seeing.

Best regards,

Paul

--
You received this message because you are subscribed to the Google Groups "OpenMC Users Group" group.
To unsubscribe from this group and stop receiving emails from it, send an email to openmc-users...@googlegroups.com.
To view this discussion on the web visit https://groups.google.com/d/msgid/openmc-users/11771d2f-9a06-49f2-8358-585f2fc949e9o%40googlegroups.com.

[Yue JIN]

Thank you, Jiankai. very helpful

在 2020年7月14日星期二 UTC+8下午9:25:23，Jiankai YU写道：