Hello good people. I have a problem with the depletion calculation. First, I have generated the chain file with openmc-make-depletion-chain , Now below is the depletion and settings section of my design I am using for depletion calculation. In the code, I have the depletion input paraments set and for power, I used the reactor thermal power value. I have defined a volume for my fuel, I believe with st_fuel.volume = area[Bundle_A_fuel] * 200 whatever a cell I defined, which is filled with st_fuel, will use this volume? But one thing I am not clear is this material volume I defined I don’t know how to use it in calculation also the sample example is not clear on this. meaning the st_fuel.volume hasn’t been used in the later section.
Now the real problem is with this depletion code, I see the simulation run for first time with the proper Keff. But the second time the Keff is very low (1/4) of the first run, and it goes down even more with the next run. Until the 3-4 time, I see an error ERROR: No fission sites banked on MPI rank 0
also before each run, I see warning of a material having negative density like
WARNING: nuclide Th227 in material 1 is negative (density = -9.760997752436385e-21 at/barn-cm)
WARNING: nuclide Th228 in material 1 is negative (density = -9.802581231893582e-21 at/barn-cm)
WARNING: nuclide Th232 in material 1 is negative (density = -1.4292281261795273e-18 at/barn-cm)
WARNING: nuclide Th233 in material 1 is negative (density = -1.7339315514675744e-20 at/barn-cm)
WARNING: nuclide Th234 in material 1 is negative (density = -2.3299574072632586e-17 at/barn-cm)
WARNING: nuclide Pa233 in material 1 is negative (density = -3.853794303347148e-21 at/barn-cm)
WARNING: nuclide U235 in material 1 is negative (density = -6.965309863028316e-20 at/barn-cm)
import openmc
import openmc.deplete
import numpy as np
import matplotlib.pyplot as plt
OpenMC simulation parameters
batches = 60
inactive = 10
particles = 1000