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took into account PR comments, got rid of redundant exports and dictionaries

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guillaume 2018-04-02 14:23:36 -04:00
parent 53da3af815
commit cade109745
1 changed files with 6 additions and 19 deletions
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25
examples/python/pincell_depletion/run_depletion.py
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@ -1,4 +1,5 @@
import openmc
import openmc.deplete
import numpy as np
###############################################################################
@ -11,9 +12,9 @@ inactive = 10
particles = 1000
# Depletion simulation parameters
time_step = 1.*24*60*60 # s
final_time = 15.*24*60*60 # s
time_steps = np.full(np.int(final_time / time_step), time_step)
time_step = 1*24*60*60 # s
final_time = 15*24*60*60 # s
time_steps = np.full(final_time // time_step, time_step)
chain_file = './chain_simple.xml'
power = 174 # W/cm, for 2D simulations only (use W for 3D)
@ -32,7 +33,6 @@ uo2.depletable = True
helium = openmc.Material(material_id=2, name='Helium for gap')
helium.set_density('g/cm3', 0.001598)
helium.add_element('He', 2.4044e-4)
helium.depletable = False
zircaloy = openmc.Material(material_id=3, name='Zircaloy 4')
zircaloy.set_density('g/cm3', 6.55)
@ -40,7 +40,6 @@ zircaloy.add_element('Sn', 0.014 , 'wo')
zircaloy.add_element('Fe', 0.00165, 'wo')
zircaloy.add_element('Cr', 0.001 , 'wo')
zircaloy.add_element('Zr', 0.98335, 'wo')
zircaloy.depletable = False
borated_water = openmc.Material(material_id=4, name='Borated water')
borated_water.set_density('g/cm3', 0.740582)
@ -48,7 +47,6 @@ borated_water.add_element('B', 4.0e-5)
borated_water.add_element('H', 5.0e-2)
borated_water.add_element('O', 2.4e-2)
borated_water.add_s_alpha_beta('c_H_in_H2O')
borated_water.depletable = False
###############################################################################
# Exporting to OpenMC geometry.xml file
@ -92,9 +90,8 @@ root = openmc.Universe(universe_id=0, name='root universe')
# Register Cells with Universe
root.add_cells([fuel, gap, clad, water])
# Instantiate a Geometry, register the root Universe, and export to XML
# Instantiate a Geometry, register the root Universe
geometry = openmc.Geometry(root)
geometry.export_to_xml()
###############################################################################
# Exporting to OpenMC materials.xml file
@ -107,10 +104,6 @@ area[fuel] = np.pi * fuel_or.coefficients['R'] ** 2
# Set materials volume for depletion. Set to an area for 2D simulations
uo2.volume = area[fuel]
# Instantiate a Materials collection and export to XML
materials_file = openmc.Materials([uo2, helium, zircaloy, borated_water])
materials_file.export_to_xml()
###############################################################################
# Exporting to OpenMC settings.xml file
###############################################################################
@ -131,14 +124,12 @@ entropy_mesh.lower_left = [-0.39218, -0.39218, -1.e50]
entropy_mesh.upper_right = [0.39218, 0.39218, 1.e50]
entropy_mesh.dimension = [10, 10, 1]
settings_file.entropy_mesh = entropy_mesh
settings_file.export_to_xml()
###############################################################################
# Initialize and run depletion calculation
###############################################################################
op = openmc.deplete.Operator(geometry, settings_file, chain_file)
op.round_number = True
# Perform simulation using the predictor algorithm
openmc.deplete.integrator.predictor(op, time_steps, power)
@ -150,12 +141,8 @@ openmc.deplete.integrator.predictor(op, time_steps, power)
# Open results file
results = openmc.deplete.ResultsList("depletion_results.h5")
# Dictionnary of materials and burned nuclides
mat_id_to_ind = results[0].mat_to_ind
nuc_to_ind = results[0].nuc_to_ind
# Obtain K_eff as a function of time
time, keff = results.get_eigenvalue()
# Obtain U235 concentration as a function of time
time, n_U235 = results.get_atoms(mat_id_to_ind['1'], nuc_to_ind['U235'])
time, n_U235 = results.get_atoms('1', 'U235')