From 109236710a9a90ecfbe3a8e5dedbc83bb0b1b1e1 Mon Sep 17 00:00:00 2001 From: guillaume Date: Sat, 24 Mar 2018 14:51:24 -0400 Subject: [PATCH] added python example for depletion --- .../python/pincell_depletion/chain_simple.xml | 49 ++++++ .../python/pincell_depletion/run_depletion.py | 161 ++++++++++++++++++ 2 files changed, 210 insertions(+) create mode 100644 examples/python/pincell_depletion/chain_simple.xml create mode 100644 examples/python/pincell_depletion/run_depletion.py diff --git a/examples/python/pincell_depletion/chain_simple.xml b/examples/python/pincell_depletion/chain_simple.xml new file mode 100644 index 000000000..c2e50a370 --- /dev/null +++ b/examples/python/pincell_depletion/chain_simple.xml @@ -0,0 +1,49 @@ + + + + + + + + + + + + + + + + + + + + + + 2.53000e-02 + + Gd157 Gd156 I135 Xe135 Xe136 Cs135 + 1.093250e-04 2.087260e-04 2.780820e-02 6.759540e-03 2.392300e-02 4.356330e-05 + + + + + + + 2.53000e-02 + + Gd157 Gd156 I135 Xe135 Xe136 Cs135 + 6.142710e-5 1.483250e-04 0.0292737 0.002566345 0.0219242 4.9097e-6 + + + + + + + 2.53000e-02 + + Gd157 Gd156 I135 Xe135 Xe136 Cs135 + 4.141120e-04 7.605360e-04 0.0135457 0.00026864 0.0024432 3.7100E-07 + + + + diff --git a/examples/python/pincell_depletion/run_depletion.py b/examples/python/pincell_depletion/run_depletion.py new file mode 100644 index 000000000..0ee2a411f --- /dev/null +++ b/examples/python/pincell_depletion/run_depletion.py @@ -0,0 +1,161 @@ +import openmc +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 = 45.*24*60*60 # s +time_steps = np.full(np.int(final_time / time_step), time_step) + +chain_file = './chain_simple.xml' +power = 174.1 # W/cm, for 2D simulations only + +############################################################################### +# Define materials +############################################################################### + +# Instantiate some Materials and register the appropriate Nuclides +uo2 = openmc.Material(material_id=1, name='UO2 fuel at 2.4% wt enrichment') +uo2.set_density('g/cm3', 10.29769) +uo2.add_element('U', 1., enrichment=2.4) +uo2.add_element('O', 2.) +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) +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) +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 +############################################################################### + +# Instantiate ZCylinder surfaces +fuel_or = openmc.ZCylinder(surface_id=1, x0=0, y0=0, R=0.39218, name='Fuel OR') +clad_ir = openmc.ZCylinder(surface_id=2, x0=0, y0=0, R=0.40005, name='Clad IR') +clad_or = openmc.ZCylinder(surface_id=3, x0=0, y0=0, R=0.45720, name='Clad OR') +left = openmc.XPlane(surface_id=4, x0=-0.62992, name='left') +right = openmc.XPlane(surface_id=5, x0=0.62992, name='right') +bottom = openmc.YPlane(surface_id=6, y0=-0.62992, name='bottom') +top = openmc.YPlane(surface_id=7, y0=0.62992, name='top') + +left.boundary_type = 'reflective' +right.boundary_type = 'reflective' +top.boundary_type = 'reflective' +bottom.boundary_type = 'reflective' + +# Instantiate Cells +fuel = openmc.Cell(cell_id=1, name='cell 1') +gap = openmc.Cell(cell_id=2, name='cell 2') +clad = openmc.Cell(cell_id=3, name='cell 3') +water = openmc.Cell(cell_id=4, name='cell 4') + +# Use surface half-spaces to define regions +fuel.region = -fuel_or +gap.region = +fuel_or & -clad_ir +clad.region = +clad_ir & -clad_or +water.region = +clad_or & +left & -right & +bottom & -top + +# Register Materials with Cells +fuel.fill = uo2 +gap.fill = helium +clad.fill = zircaloy +water.fill = borated_water + +# Instantiate Universe +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 +geometry = openmc.Geometry(root) +geometry.export_to_xml() + +############################################################################### +# Exporting to OpenMC materials.xml file +############################################################################### + +# Compute cell areas +area = {} +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 +############################################################################### + +# 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 +settings_file.export_to_xml() + +############################################################################### +# 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) + +############################################################################### +# Read depletion calculation results +############################################################################### + +# 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, U5 = results.get_atoms(mat_id_to_ind['1'], nuc_to_ind['U235'])