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Merge pull request #987 from GiudGiud/develop
Added a python example for depletion
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Paul Romano
GitHub
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49
examples/python/pincell_depletion/chain_simple.xml
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49
examples/python/pincell_depletion/chain_simple.xml
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<?xml version="1.0"?>
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<depletion_chain>
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<nuclide name="I135" decay_modes="1" reactions="1" half_life="2.36520E+04">
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<decay type="beta" target="Xe135" branching_ratio="1.0" />
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<reaction type="(n,gamma)" Q="0.0" target="Xe136" /> <!-- Not precisely true, but whatever -->
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</nuclide>
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<nuclide name="Xe135" decay_modes="1" reactions="1" half_life="3.29040E+04">
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<decay type=" beta" target="Cs135" branching_ratio="1.0" />
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<reaction type="(n,gamma)" Q="0.0" target="Xe136" />
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</nuclide>
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<nuclide name="Xe136" decay_modes="0" reactions="0" />
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<nuclide name="Cs135" decay_modes="0" reactions="0" />
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<nuclide name="Gd157" decay_modes="0" reactions="1" >
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<reaction type="(n,gamma)" Q="0.0" target="Nothing" />
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</nuclide>
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<nuclide name="Gd156" decay_modes="0" reactions="1">
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<reaction type="(n,gamma)" Q="0.0" target="Gd157" />
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</nuclide>
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<nuclide name="U234" decay_modes="0" reactions="1">
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<reaction type="fission" Q="191840000."/>
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<neutron_fission_yields>
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<energies>2.53000e-02</energies>
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<fission_yields energy="2.53000e-02">
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<products>Gd157 Gd156 I135 Xe135 Xe136 Cs135</products>
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<data>1.093250e-04 2.087260e-04 2.780820e-02 6.759540e-03 2.392300e-02 4.356330e-05</data>
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</fission_yields>
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</neutron_fission_yields>
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</nuclide>
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<nuclide name="U235" decay_modes="0" reactions="1">
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<reaction type="fission" Q="193410000."/>
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<neutron_fission_yields>
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<energies>2.53000e-02</energies>
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<fission_yields energy="2.53000e-02">
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<products>Gd157 Gd156 I135 Xe135 Xe136 Cs135</products>
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<data>6.142710e-5 1.483250e-04 0.0292737 0.002566345 0.0219242 4.9097e-6</data>
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</fission_yields>
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</neutron_fission_yields>
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</nuclide>
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<nuclide name="U238" decay_modes="0" reactions="1">
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<reaction type="fission" Q="197790000."/>
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<neutron_fission_yields>
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<energies>2.53000e-02</energies>
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<fission_yields energy="2.53000e-02">
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<products>Gd157 Gd156 I135 Xe135 Xe136 Cs135</products>
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<data>4.141120e-04 7.605360e-04 0.0135457 0.00026864 0.0024432 3.7100E-07</data>
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</fission_yields>
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</neutron_fission_yields>
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</nuclide>
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</depletion_chain>
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148
examples/python/pincell_depletion/run_depletion.py
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examples/python/pincell_depletion/run_depletion.py
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import openmc
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import openmc.deplete
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import numpy as np
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###############################################################################
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# Simulation Input File Parameters
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###############################################################################
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# OpenMC simulation parameters
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batches = 100
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inactive = 10
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particles = 1000
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# Depletion simulation parameters
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time_step = 1*24*60*60 # s
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final_time = 15*24*60*60 # s
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time_steps = np.full(final_time // time_step, time_step)
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chain_file = './chain_simple.xml'
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power = 174 # W/cm, for 2D simulations only (use W for 3D)
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###############################################################################
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# Define materials
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###############################################################################
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# Instantiate some Materials and register the appropriate Nuclides
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uo2 = openmc.Material(material_id=1, name='UO2 fuel at 2.4% wt enrichment')
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uo2.set_density('g/cm3', 10.29769)
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uo2.add_element('U', 1., enrichment=2.4)
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uo2.add_element('O', 2.)
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uo2.depletable = True
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helium = openmc.Material(material_id=2, name='Helium for gap')
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helium.set_density('g/cm3', 0.001598)
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helium.add_element('He', 2.4044e-4)
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zircaloy = openmc.Material(material_id=3, name='Zircaloy 4')
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zircaloy.set_density('g/cm3', 6.55)
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zircaloy.add_element('Sn', 0.014 , 'wo')
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zircaloy.add_element('Fe', 0.00165, 'wo')
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zircaloy.add_element('Cr', 0.001 , 'wo')
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zircaloy.add_element('Zr', 0.98335, 'wo')
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borated_water = openmc.Material(material_id=4, name='Borated water')
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borated_water.set_density('g/cm3', 0.740582)
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borated_water.add_element('B', 4.0e-5)
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borated_water.add_element('H', 5.0e-2)
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borated_water.add_element('O', 2.4e-2)
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borated_water.add_s_alpha_beta('c_H_in_H2O')
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###############################################################################
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# Exporting to OpenMC geometry.xml file
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###############################################################################
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# Instantiate ZCylinder surfaces
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fuel_or = openmc.ZCylinder(surface_id=1, x0=0, y0=0, R=0.39218, name='Fuel OR')
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clad_ir = openmc.ZCylinder(surface_id=2, x0=0, y0=0, R=0.40005, name='Clad IR')
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clad_or = openmc.ZCylinder(surface_id=3, x0=0, y0=0, R=0.45720, name='Clad OR')
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left = openmc.XPlane(surface_id=4, x0=-0.62992, name='left')
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right = openmc.XPlane(surface_id=5, x0=0.62992, name='right')
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bottom = openmc.YPlane(surface_id=6, y0=-0.62992, name='bottom')
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top = openmc.YPlane(surface_id=7, y0=0.62992, name='top')
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left.boundary_type = 'reflective'
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right.boundary_type = 'reflective'
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top.boundary_type = 'reflective'
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bottom.boundary_type = 'reflective'
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# Instantiate Cells
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fuel = openmc.Cell(cell_id=1, name='cell 1')
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gap = openmc.Cell(cell_id=2, name='cell 2')
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clad = openmc.Cell(cell_id=3, name='cell 3')
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water = openmc.Cell(cell_id=4, name='cell 4')
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# Use surface half-spaces to define regions
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fuel.region = -fuel_or
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gap.region = +fuel_or & -clad_ir
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clad.region = +clad_ir & -clad_or
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water.region = +clad_or & +left & -right & +bottom & -top
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# Register Materials with Cells
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fuel.fill = uo2
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gap.fill = helium
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clad.fill = zircaloy
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water.fill = borated_water
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# Instantiate Universe
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root = openmc.Universe(universe_id=0, name='root universe')
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# Register Cells with Universe
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root.add_cells([fuel, gap, clad, water])
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# Instantiate a Geometry, register the root Universe
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geometry = openmc.Geometry(root)
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###############################################################################
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# Exporting to OpenMC materials.xml file
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###############################################################################
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# Compute cell areas
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area = {}
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area[fuel] = np.pi * fuel_or.coefficients['R'] ** 2
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# Set materials volume for depletion. Set to an area for 2D simulations
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uo2.volume = area[fuel]
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###############################################################################
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# Exporting to OpenMC settings.xml file
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###############################################################################
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# Instantiate a Settings object, set all runtime parameters, and export to XML
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settings_file = openmc.Settings()
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settings_file.batches = batches
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settings_file.inactive = inactive
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settings_file.particles = particles
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# Create an initial uniform spatial source distribution over fissionable zones
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bounds = [-0.62992, -0.62992, -1, 0.62992, 0.62992, 1]
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uniform_dist = openmc.stats.Box(bounds[:3], bounds[3:], only_fissionable=True)
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settings_file.source = openmc.source.Source(space=uniform_dist)
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entropy_mesh = openmc.Mesh()
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entropy_mesh.lower_left = [-0.39218, -0.39218, -1.e50]
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entropy_mesh.upper_right = [0.39218, 0.39218, 1.e50]
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entropy_mesh.dimension = [10, 10, 1]
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settings_file.entropy_mesh = entropy_mesh
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###############################################################################
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# Initialize and run depletion calculation
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###############################################################################
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op = openmc.deplete.Operator(geometry, settings_file, chain_file)
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# Perform simulation using the predictor algorithm
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openmc.deplete.integrator.predictor(op, time_steps, power)
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###############################################################################
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# Read depletion calculation results
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###############################################################################
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# Open results file
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results = openmc.deplete.ResultsList("depletion_results.h5")
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# Obtain K_eff as a function of time
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time, keff = results.get_eigenvalue()
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# Obtain U235 concentration as a function of time
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time, n_U235 = results.get_atoms('1', 'U235')
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