Full-core depletion

User Support
1

Hi,

I’m trying to extend the depletion calculation from a single assembly to a full reactor core model. To test the model before run a long- time simulation. I set up a calculation with 9 hours of the reactor operation time, and the step size is an hour. For the transport calculation, the number of batches is 250 and every batch have 10000 particle history.

The results are quite interesting. The k-eff values drop down very soon which would be presented in the plot. I believe that because of Xe gas, maybe that is reasonable.
k-eff
However, the output file also has some warnings about the nuclide number, looks like:

warning
warning1631×624 85.4 KB

The negative values of nuclide number look like not normal, I am worried this effect the calculation results. If anyone knows the reason why the results look like that please let me know.

Best,
Yiming

2

Yes, for a thermal spectrum reactor, the buildup of xenon will quickly decrease keff when depleting. As far as the negative number densities, note that the values are very low (on the order of 1e-20 atom/b-cm) and are effectively zero. Some things you may want to consider that could cause this:

  • Is your power specified in the correct units? (OpenMC expects W, not MW or something else)
  • Timestep lengths that are too long can also cause issues, but that shouldn’t be a problem in your case where the steps are quite short.
3

Hi Paul,

Thanks for your reply. It’s good to hear the negative number densities will not affect the results. The power I specified in this simulation is 330e6, which unit should be Watt, not MW or something else.

By the way, may I ask another question? Due to the application of ‘diff_burnable_mats’ argument, there are thousands of ‘fuel’ materials in the material.xml file. But it brings some challenge to the analysis of the results. The first required parameter in the openmc.deplete.ReactionRates [Python API is the Material ID of local_mats(list of str). Is there any method to get a particular reaction rate or a specified atom number in all the ‘fuel’ burnable materials at the same time?

Many thanks for your help.

Yiming

4

@YimingZ There is no automated way to get the reaction rate of atom density in all materials, but it is fairly easy to get at that information in Python. The following would give you a list of all the material IDs (as strings) that you can use in subsequent calls to get_atoms or get_reaction_rate:

# Get results list object
results = openmc.deplete.ResultsList.from_hdf5('depletion_results.h5')

# Get list of material IDs
first_result = results[0]
material_ids = list(first_result.mat_to_ind.keys())

# Example: get U235 density for all materials
u235_density = [results.get_atoms(i, "U235") for i in material_ids]