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https://github.com/openmc-dev/openmc.git
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Added tests of the openmc.Model class. Still need to do tests of deplete and the methods which change python and C attributes. Pushing this now to verify the CI shows passes for the MPI configs
This commit is contained in:
agnelson
parent
24bca06310
commit
0d2f923be8
3 changed files with 308 additions and 3 deletions
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@ -193,6 +193,8 @@ class Model:
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def from_xml(cls, geometry='geometry.xml', materials='materials.xml',
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settings='settings.xml'):
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"""Create model from existing XML files
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When initializing this way, the user must manually load plots, tallies,
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the chain_file and fission_q attributes.
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Parameters
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----------
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@ -428,7 +430,9 @@ class Model:
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Parameters
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----------
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**kwargs
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Keyword arguments passed to :func:`openmc.run`
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Keyword arguments passed to :func:`openmc.run`. Note that these are
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ignored if running via the C-API. Instead the parameters should be
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set via the Settings object.
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Returns
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-------
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@ -1,9 +1,129 @@
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from math import pi
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import openmc
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import pytest
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from tests.regression_tests import config
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@pytest.fixture(scope='module')
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def pin_model_attributes():
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uo2 = openmc.Material(name='UO2')
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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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zirc = openmc.Material(name='Zirc')
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zirc.set_density('g/cm3', 6.55)
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zirc.add_element('Zr', 1.)
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borated_water = openmc.Material(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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mats = openmc.Materials([uo2, zirc, borated_water])
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pitch = 1.25984
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fuel_or = openmc.ZCylinder(r=0.39218, name='Fuel OR')
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clad_or = openmc.ZCylinder(r=0.45720, name='Clad OR')
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box = openmc.model.rectangular_prism(pitch, pitch, boundary_type='reflective')
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# Define cells
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fuel = openmc.Cell(name='fuel', fill=uo2, region=-fuel_or)
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clad = openmc.Cell(fill=zirc, region=+fuel_or & -clad_or)
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water = openmc.Cell(fill=borated_water, region=+clad_or & box)
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# Define overall geometry
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geom = openmc.Geometry([fuel, clad, water])
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uo2.volume = pi * fuel_or.r**2
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settings = openmc.Settings()
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settings.batches = 100
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settings.inactive = 10
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settings.particles = 1000
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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.source = openmc.source.Source(space=uniform_dist)
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entropy_mesh = openmc.RegularMesh()
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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.entropy_mesh = entropy_mesh
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tals = openmc.Tallies()
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tal = openmc.Tally(name='test')
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tal.scores = ['flux']
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tals.append(tal)
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plot = openmc.Plot()
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plot.origin = (0., 0., 0.)
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plot.width = (pitch, pitch)
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plot.pixels = (300, 300)
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plot.color_by = 'material'
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plots = openmc.Plots((plot,))
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chain = './chain_simple.xml'
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fission_q = {'U235': 200e6}
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chain_file_xml = """<?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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"""
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return (mats, geom, settings, tals, plots, chain, fission_q,
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chain_file_xml)
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@pytest.fixture(scope='module')
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def mpi_intracomm():
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if config['mpi']:
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@ -1,18 +1,162 @@
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from typing import Type
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import pytest
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from pathlib import Path
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import openmc
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import openmc.lib
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def test_init(run_in_tmpdir, pin_model_attributes):
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mats, geom, settings, tals, plots, chain, fission_q, _ = \
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pin_model_attributes
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openmc.reset_auto_ids()
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# Check blank initialization of a model
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test_model = openmc.Model()
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assert test_model.geometry.root_universe is None
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assert len(test_model.materials) == 0
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ref_settings = openmc.Settings()
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assert sorted(test_model.settings.__dict__.keys()) == \
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sorted(ref_settings.__dict__.keys())
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for ref_k, ref_v in ref_settings.__dict__.items():
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assert test_model.settings.__dict__[ref_k] == ref_v
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assert len(test_model.tallies) == 0
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assert len(test_model.plots) == 0
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assert test_model.chain_file is None
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assert test_model.fission_q is None
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assert test_model._materials_by_id == {}
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assert test_model._materials_by_name == {}
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assert test_model._cells_by_id == {}
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assert test_model._cells_by_name == {}
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assert test_model.C_init is False
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# Now check proper init of an actual model. Assume no interference between
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# parameters and so we can apply them all at once instead of testing one
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# parameter initialization at a time
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test_model = openmc.Model(geom, mats, settings, tals, plots, chain,
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fission_q)
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assert test_model.geometry is geom
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assert test_model.materials is mats
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assert test_model.settings is settings
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assert test_model.tallies is tals
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assert test_model.plots is plots
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assert test_model.chain_file == Path(chain).resolve()
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assert test_model.fission_q is fission_q
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assert test_model._materials_by_id == {1: mats[0], 2: mats[1], 3: mats[2]}
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assert test_model._materials_by_name == {'UO2': mats[0], 'Zirc': mats[1],
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'Borated water': mats[2]}
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assert test_model._cells_by_id == {1: geom.root_universe.cells[1],
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2: geom.root_universe.cells[2],
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3: geom.root_universe.cells[3]}
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# No cell name for 2 and 3, so we expect a blank name to be assigned to
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# cell 3 due to overwriting
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assert test_model._cells_by_name == {
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'fuel': geom.root_universe.cells[1], '': geom.root_universe.cells[3]}
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assert test_model.C_init is False
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# Finally test the parameter type checking by passing bad types and
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# obtaining the right exception types
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def_params = [geom, mats, settings, tals, plots, chain, fission_q]
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for i in range(len(def_params)):
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args = def_params.copy()
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# Try an integer, as that is a bad type for all arguments
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args[i] = i
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with pytest.raises(TypeError):
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test_model = openmc.Model(*args)
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def test_from_xml(run_in_tmpdir, pin_model_attributes):
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mats, geom, settings, tals, plots, _, _, _ = pin_model_attributes
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# This test will write the individual files to xml and then init that way
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# and run the same sort of test as in test_init
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mats.export_to_xml()
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geom.export_to_xml()
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settings.export_to_xml()
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tals.export_to_xml()
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plots.export_to_xml()
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# This from_xml method cannot load chain and fission_q
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test_model = openmc.Model.from_xml()
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assert test_model.geometry.root_universe.cells.keys() == \
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geom.root_universe.cells.keys()
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assert [c.fill.name for c in test_model.geometry.root_universe.cells.values()] == \
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[c.fill.name for c in geom.root_universe.cells.values()]
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assert [mat.name for mat in test_model.materials] == [mat.name for mat in mats]
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# We will assume the attributes of settings that are custom objects are
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# OK if the others are so we dotn need to implement explicit comparisons
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no_test = ['_source', '_entropy_mesh']
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assert sorted(k for k in test_model.settings.__dict__.keys() if k not in no_test) == \
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sorted(k for k in settings.__dict__.keys() if k not in no_test)
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keys = sorted(k for k in settings.__dict__.keys() if k not in no_test)
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for ref_k in keys:
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assert test_model.settings.__dict__[ref_k] == settings.__dict__[ref_k]
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assert len(test_model.tallies) == 0
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assert len(test_model.plots) == 0
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assert test_model.chain_file is None
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assert test_model.fission_q is None
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assert test_model._materials_by_id == \
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{1: test_model.materials[0], 2: test_model.materials[1],
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3: test_model.materials[2]}
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assert test_model._materials_by_name == {
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'UO2': test_model.materials[0], 'Zirc': test_model.materials[1],
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'Borated water': test_model.materials[2]}
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assert test_model._cells_by_id == {\
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1: test_model.geometry.root_universe.cells[1],
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2: test_model.geometry.root_universe.cells[2],
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3: test_model.geometry.root_universe.cells[3]}
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# No cell name for 2 and 3, so we expect a blank name to be assigned to
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# cell 3 due to overwriting
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assert test_model._cells_by_name == {
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'fuel': test_model.geometry.root_universe.cells[1],
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'': test_model.geometry.root_universe.cells[3]}
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assert test_model.C_init is False
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def test_init_clear_C_api(run_in_tmpdir, pin_model_attributes, mpi_intracomm):
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# We are going to init and then make sure data is loaded
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mats, geom, settings, tals, plots, _, _, _ = pin_model_attributes
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test_model = openmc.Model(geom, mats, settings, tals, plots)
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# Set the depletion module to use the test intracomm as the depletion mods
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# intracomm is the one that init uses. We will not perturb system state
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# outside of this test by re-setting the openmc.deplete.comm to what it was
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# before when we exist
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if mpi_intracomm is not None:
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orig_comm = openmc.deplete.comm
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openmc.deplete.comm = mpi_intracomm
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test_model.init_C_api()
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# First check that the API is advertised as initialized
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assert openmc.lib.LIB_INIT is True
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assert test_model.C_init is True
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# Now make sure it actually is initialized by making a call to the lib
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c_mat = openmc.lib.find_material((0.6, 0., 0.))
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# This should be Borated water
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assert c_mat.name == 'Borated water'
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assert c_mat.id == 3
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# Ok, now lets test that we can clear the data and check that it is cleared
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test_model.clear_C_api()
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# First check that the API is advertised as initialized
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assert openmc.lib.LIB_INIT is False
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assert test_model.C_init is False
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# Note we cant actually test that a sys call fails because we should get a
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# seg fault
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# And before done, reset the deplete communicator
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if mpi_intracomm is not None:
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openmc.deplete.comm = orig_comm
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def test_import_properties(run_in_tmpdir, mpi_intracomm):
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"""Test importing properties on the Model class """
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# Create PWR pin cell model and write XML files
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openmc.reset_auto_ids()
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model = openmc.examples.pwr_pin_cell()
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model.export_to_xml()
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model.init_C_api()
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# Change fuel temperature and density and export properties
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openmc.lib.init(intracomm=mpi_intracomm)
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cell = openmc.lib.cells[1]
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cell.set_temperature(600.0)
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cell.fill.set_density(5.0, 'g/cm3')
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@ -32,3 +176,40 @@ def test_import_properties(run_in_tmpdir, mpi_intracomm):
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cell = model_with_properties.geometry.get_all_cells()[1]
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assert cell.temperature == [600.0]
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assert cell.fill.get_mass_density() == pytest.approx(5.0)
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def test_run(run_in_tmpdir, pin_model_attributes, mpi_intracomm):
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mats, geom, settings, tals, plots, _, _, _ = pin_model_attributes
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test_model = openmc.Model(geom, mats, settings, tals, plots)
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# Set the depletion module to use the test intracomm as the depletion mods
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# intracomm is the one that init uses. We will not perturb system state
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# outside of this test by re-setting the openmc.deplete.comm to what it was
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# before when we exist
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# This case will run by getting the k-eff and tallies for command-line and
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# C-API execution modes and ensuring they give the same result.
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sp_path = test_model.run(output=False)
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with openmc.StatePoint(sp_path) as sp:
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cli_keff = sp.k_combined
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cli_flux = sp.get_tally(id=1).get_values()[0, 0, 0]
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if mpi_intracomm is not None:
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orig_comm = openmc.deplete.comm
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openmc.deplete.comm = mpi_intracomm
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test_model.settings.verbosity = 1
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test_model.init_C_api()
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sp_path = test_model.run()
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with openmc.StatePoint(sp_path) as sp:
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C_keff = sp.k_combined
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C_flux = sp.get_tally(id=1).get_values()[0, 0, 0]
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test_model.clear_C_api()
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# and lets compare results
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assert (C_keff - cli_keff) < 1e-15
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assert (C_flux - cli_flux) < 1e-15
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# And before done, reset the deplete communicator
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if mpi_intracomm is not None:
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openmc.deplete.comm = orig_comm
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