Weird Fission Reaction Rate Calculated in Epithermal Assembly Depletion

Hello all,

I have run into a weird problem while doing Epithermal Assembly Depletion Multiphysics Simulation using OPENMC. I Compared the Fission Reaction Rate (Reaction/s) from OPENMC with the existing one from SERPENT. As I had reproduced and recheck the assembly many times and was quite sure that nothing was wrong in the geometry and material Volume / Composition / Temperature / Density. I think maybe there are some things I don’t know inside OPENMC that caused this. (Maybe the Normalization Method ? or The different unresolved resonance treatments ? )

The result from OPENMC (Left) and SERPENT (right) -

  1. The Magnitude of the OPENMC’s result is about to double the SERPENT result
  2. The Trend seems similar but there’s some fission rate growing in the internal blank zone ( 22.1 – 78.1 cm) which the Burnable fuel Mat contains U235 U238 and O16.

after I extracted the reaction rate of each nuclei , I found out that the reaction rate of PU239 is the main variable causing this overrate and reshape the overall trend in the internal blanket zone

I plot the result again without PU239 and U238 (still keeping U235 in the summation) the trend and magnitude are quite almost similar to each other.

Note 1 : I got the total reaction rate in OPENMC by running :

FissionXXX_Temp = [results.get_reaction_rate(i, “XXX”, “fission”) for i in material_ids]

for the important nuclei existed including

U232,U233,U234,U235,U236,U237, U238, U239
Np238, Np236, Np237, Np239
Pu243, Pu236, Pu237, Pu238, Pu239, Pu240, Pu241, Pu242, 
Am241, Am242, Am243
Cm243, Cm244, Cm245, Cm246, Cm242
Pa231, Pa232, Pa233
Th230, Th231, Th232, Th233, Th234

Then summation all the reaction rates.

Note 2 : here’s my integrator setting :

chain = openmc.deplete.Chain.from_xml("./chain_casl_pwr.xml")
# Power Density (kW/gm) # set powdens 5.207e-2
operator = openmc.deplete.Operator(stacktest0, settings_file,"./chain_casl_pwr.xml", prev_results=r)
integrator = openmc.deplete.PredictorIntegrator(operator,time_steps_Now,power_density = powdens, timestep_units='MWd/kg')

Note 3 : here’s my XML setting input :

# Instantiate a Settings object

settings_file = openmc.Settings()
settings_file.batches = batches  #250
settings_file.inactive = inactive #50
settings_file.particles = particles #150000
settings_file.output = {'tallies': True}
settings_file.run_mode = 'eigenvalue'
settings_file.ptables = True
settings_file.resonance_scattering = {"enable":True, "method":"dbrc"}
settings_file.temperature['interpolation'] = True
bounds = [-0.63, -0.63, -0.63, 120, 120, 120]
uniform_dist = openmc.stats.Box(bounds[:3], bounds[3:], only_fissionable=True)
settings_file.source = openmc.Source(space=uniform_dist)

Thank you forward for any suggestion or Explanation about this problem
PS. Anyway i had no problem with the decaying of Keff

1 Like

I finally solved the problem

Hi @Sivakorn.SSW ,

Nice to hear that you solve it finally!
Just for curiosity, how did you that? :yum:


Hi @bentridisalah ,

There’s some difference in using [MeshFilter and CellFilter ] I got the result which are similar to serpent by using MeshFilter, However, I would prefer to use the CellFilter as I’m doing Multiphysics calculation loop and changing reactor block height, dealing with Meshfilter will be really complicated. so It’s just a temporary solution.

Here’s what I found : Difference in Fission rate Tally value from MeshFilter and CellFilter

thank you very much @Sivakorn.SSW. It should help me a lot for my calculations on both topcis: Oklo natural nuclear reactor and Th-U SMR modelling

keep in touch