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68 lines
2 KiB
Python
68 lines
2 KiB
Python
"""Test one-group cross section generation"""
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from pathlib import Path
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import numpy as np
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import pytest
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import openmc
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from openmc.deplete import MicroXS, get_microxs_and_flux
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from tests.regression_tests import config
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CHAIN_FILE = Path(__file__).parents[2] / "chain_simple.xml"
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@pytest.fixture(scope="module")
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def model():
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fuel = openmc.Material(name="uo2")
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fuel.add_nuclide("U235", 1.0)
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fuel.add_nuclide("O16", 2.0)
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fuel.set_density("g/cc", 10.4)
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sphere = openmc.Sphere(r=10.0, boundary_type='vacuum')
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cell = openmc.Cell(region=-sphere, fill=fuel)
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geometry = openmc.Geometry([cell])
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settings = openmc.Settings()
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settings.particles = 1000
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settings.inactive = 5
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settings.batches = 10
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return openmc.Model(geometry, settings=settings)
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@pytest.mark.parametrize(
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"domain_type, rr_mode",
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[
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("materials", "direct"),
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("materials", "flux"),
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("mesh", "direct"),
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("mesh", "flux"),
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]
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)
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def test_from_model(model, domain_type, rr_mode):
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if domain_type == 'materials':
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domains = list(model.geometry.get_all_materials().values())
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elif domain_type == 'mesh':
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mesh = openmc.RegularMesh()
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mesh.lower_left = (-10., -10.)
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mesh.upper_right = (10., 10.)
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mesh.dimension = (1, 1)
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domains = mesh
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nuclides = ['U235', 'O16', 'Xe135']
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kwargs = {
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'reaction_rate_mode': rr_mode,
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'chain_file': CHAIN_FILE,
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'path_statepoint': 'neutron_transport.h5',
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}
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if rr_mode == 'flux':
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kwargs['energies'] = 'CASMO-40'
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_, test_xs = get_microxs_and_flux(model, domains, nuclides, **kwargs)
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if config['update']:
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test_xs[0].to_csv(f'test_reference_{domain_type}_{rr_mode}.csv')
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# Make sure results match reference results
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ref_xs = MicroXS.from_csv(f'test_reference_{domain_type}_{rr_mode}.csv')
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np.testing.assert_allclose(test_xs[0].data, ref_xs.data, rtol=1e-11)
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# Make sure statepoint file was saved
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assert Path('neutron_transport.h5').exists()
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Path('neutron_transport.h5').unlink()
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