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added depletion restart python example

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guillaume 2018-05-28 18:38:31 -04:00
parent 51911bacb4
commit ed65c15b30
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examples/python/pincell_depletion/restart_depletion.py Normal file
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import openmc
import openmc.deplete
import numpy as np
###############################################################################
# Simulation Input File Parameters
###############################################################################
# OpenMC simulation parameters
batches = 100
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(final_time // time_step, time_step)
chain_file = './chain_simple.xml'
power = 174 # W/cm, for 2D simulations only (use W for 3D)
###############################################################################
# Load previous simulation results
###############################################################################
# Load geometry from statepoint
statepoint = 'statepoint.100.h5'
sp = openmc.StatePoint(statepoint)
geometry = sp.summary.geometry
# Load previous delpletion results
previous_results = openmc.deplete.ResultsList("depletion_results.h5")
# Reload volumes into geometry
previous_results.transfer_volumes(geometry)
###############################################################################
# Set transport calculation settings
###############################################################################
# Instantiate a Settings object, set all runtime parameters, and export to XML
settings_file = openmc.Settings()
settings_file.batches = batches
settings_file.inactive = inactive
settings_file.particles = particles
# Create an initial uniform spatial source distribution over fissionable zones
bounds = [-0.62992, -0.62992, -1, 0.62992, 0.62992, 1]
uniform_dist = openmc.stats.Box(bounds[:3], bounds[3:], only_fissionable=True)
settings_file.source = openmc.source.Source(space=uniform_dist)
entropy_mesh = openmc.Mesh()
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
###############################################################################
# Initialize and run depletion calculation
###############################################################################
op = openmc.deplete.Operator(geometry, settings_file, chain_file)
# Perform simulation using the predictor algorithm
openmc.deplete.integrator.predictor(op, time_steps, power)
###############################################################################
# Read depletion calculation results
###############################################################################
# Open results file
results = openmc.deplete.ResultsList("depletion_results.h5")
# 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('1', 'U235')