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https://github.com/openmc-dev/openmc.git
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New regression test for lattice with OX orient
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dr.yuri
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4 changed files with 299 additions and 0 deletions
0
tests/regression_tests/lattice_hex_OX/__init__.py
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tests/regression_tests/lattice_hex_OX/__init__.py
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126
tests/regression_tests/lattice_hex_OX/inputs_true.dat
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tests/regression_tests/lattice_hex_OX/inputs_true.dat
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<?xml version='1.0' encoding='utf-8'?>
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<geometry>
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<cell id="1" material="1" region="-1" universe="1" />
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<cell id="2" material="4" region="-2 1" universe="1" />
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<cell id="3" material="2" region="2" universe="1" />
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<cell id="4" material="2" region="-3" universe="2" />
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<cell id="5" material="4" region="-4 3" universe="2" />
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<cell id="6" material="2" region="4" universe="2" />
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<cell id="7" material="3" region="-5" universe="3" />
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<cell id="8" material="4" region="-6 5" universe="3" />
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<cell id="9" material="2" region="-7 6" universe="3" />
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<cell id="10" material="4" region="-8 7" universe="3" />
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<cell id="11" material="2" region="8" universe="3" />
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<cell id="12" material="2" universe="4" />
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<cell fill="9" id="13" name="container assembly cell" region="-11 12 -13 14 15 -16 9 -10" universe="5" />
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<hex_lattice id="9" n_axial="2" n_rings="11" name="regular fuel assembly" orient="OX">
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<pitch>1.235 5.0</pitch>
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<outer>4</outer>
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<center>0.0 0.0 5.0</center>
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<universes>
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1 1 1 1 1 1 1 1 1 1 1
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1 1 1 1 1 1 1 1 1 1 1 1
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1 1 1 1 1 1 1 1 1 1 1 1 1
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1 1 1 1 1 1 1 1 1 1 1 1 1 1
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1 1 1 1 1 1 1 3 1 1 1 1 1 1 1
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1 1 1 1 1 3 1 1 1 1 3 1 1 1 1 1
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1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1
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1 1 1 1 3 1 1 1 1 3 1 1 1 3 1 1 1 1
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1 1 1 1 1 1 1 3 1 1 1 1 1 1 1 1 1 1 1
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1 1 1 1 1 1 1 1 1 1 1 1 3 1 1 1 1 1 1 1
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1 1 1 1 1 3 1 1 1 1 2 1 1 1 1 3 1 1 1 1 1
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1 1 1 1 1 1 1 3 1 1 1 1 1 1 1 1 1 1 1 1
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1 1 1 1 1 1 1 1 1 1 1 3 1 1 1 1 1 1 1
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1 1 1 1 3 1 1 1 3 1 1 1 1 3 1 1 1 1
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1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1
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1 1 1 1 1 3 1 1 1 1 3 1 1 1 1 1
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1 1 1 1 1 1 1 3 1 1 1 1 1 1 1
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1 1 1 1 1 1 1 1 1 1 1 1 1 1
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1 1 1 1 1 1 1 1 1 1 1 1 1
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1 1 1 1 1 1 1 1 1 1 1 1
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1 1 1 1 1 1 1 1 1 1 1
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1 1 1 1 1 1 1 1 1 1 1
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1 1 1 1 1 1 1 1 1 1 1 1
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1 1 1 1 1 1 1 1 1 1 1 1 1
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1 1 1 1 1 1 1 1 1 1 1 1 1 1
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1 1 1 1 1 1 1 3 1 1 1 1 1 1 1
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1 1 1 1 1 3 1 1 1 1 3 1 1 1 1 1
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1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1
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1 1 1 1 3 1 1 1 1 3 1 1 1 3 1 1 1 1
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1 1 1 1 1 1 1 3 1 1 1 1 1 1 1 1 1 1 1
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1 1 1 1 1 1 1 1 1 1 1 1 3 1 1 1 1 1 1 1
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1 1 1 1 1 3 1 1 1 1 2 1 1 1 1 3 1 1 1 1 1
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1 1 1 1 1 1 1 3 1 1 1 1 1 1 1 1 1 1 1 1
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1 1 1 1 1 1 1 1 1 1 1 3 1 1 1 1 1 1 1
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1 1 1 1 3 1 1 1 3 1 1 1 1 3 1 1 1 1
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1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1
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1 1 1 1 1 3 1 1 1 1 3 1 1 1 1 1
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1 1 1 1 1 1 1 3 1 1 1 1 1 1 1
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1 1 1 1 1 1 1 1 1 1 1 1 1 1
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1 1 1 1 1 1 1 1 1 1 1 1 1
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1 1 1 1 1 1 1 1 1 1 1 1
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1 1 1 1 1 1 1 1 1 1 1</universes>
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</hex_lattice>
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<surface coeffs="0.0 0.0 0.386" id="1" type="z-cylinder" />
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<surface coeffs="0.0 0.0 0.4582" id="2" type="z-cylinder" />
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<surface coeffs="0.0 0.0 0.45" id="3" type="z-cylinder" />
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<surface coeffs="0.0 0.0 0.5177" id="4" type="z-cylinder" />
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<surface coeffs="0.0 0.0 0.35" id="5" type="z-cylinder" />
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<surface coeffs="0.0 0.0 0.41" id="6" type="z-cylinder" />
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<surface coeffs="0.0 0.0 0.545" id="7" type="z-cylinder" />
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<surface coeffs="0.0 0.0 0.6323" id="8" type="z-cylinder" />
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<surface boundary="reflective" coeffs="0.0" id="9" type="z-plane" />
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<surface boundary="reflective" coeffs="10.0" id="10" type="z-plane" />
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<surface boundary="reflective" coeffs="11.8" id="11" type="y-plane" />
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<surface boundary="reflective" coeffs="-11.8" id="12" type="y-plane" />
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<surface boundary="reflective" coeffs="1.7320508075688772 1.0 0.0 23.6" id="13" type="plane" />
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<surface boundary="reflective" coeffs="-1.7320508075688772 1.0 0.0 -23.6" id="14" type="plane" />
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<surface boundary="reflective" coeffs="1.7320508075688772 1.0 0.0 -23.6" id="15" type="plane" />
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<surface boundary="reflective" coeffs="-1.7320508075688772 1.0 0.0 23.6" id="16" type="plane" />
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</geometry>
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<?xml version='1.0' encoding='utf-8'?>
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<materials>
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<material depletable="true" id="1" name="UO2">
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<density units="sum" />
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<nuclide ao="0.0008737" name="U235" />
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<nuclide ao="0.018744" name="U238" />
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<nuclide ao="0.039235" name="O16" />
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</material>
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<material id="2" name="borated H2O">
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<density units="sum" />
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<nuclide ao="0.06694" name="H1" />
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<nuclide ao="0.03347" name="O16" />
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<nuclide ao="6.6262e-06" name="B10" />
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<nuclide ao="2.6839e-05" name="B11" />
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</material>
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<material id="3" name="pellet B4C">
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<density units="sum" />
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<nuclide ao="0.01966" name="C0" />
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<nuclide ao="4.7344e-06" name="B11" />
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<nuclide ao="1.9177e-05" name="B10" />
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</material>
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<material id="4" name="Zirc4">
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<density units="sum" />
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<nuclide ao="0.021763349999999997" name="Zr90" />
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<nuclide ao="0.00474606" name="Zr91" />
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<nuclide ao="0.00725445" name="Zr92" />
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<nuclide ao="0.00735174" name="Zr94" />
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<nuclide ao="0.0011844" name="Zr96" />
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</material>
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</materials>
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<?xml version='1.0' encoding='utf-8'?>
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<settings>
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<run_mode>eigenvalue</run_mode>
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<particles>1000</particles>
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<batches>10</batches>
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<inactive>5</inactive>
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<source strength="1.0">
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<space type="box">
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<parameters>-13.62546635287517 -13.62546635287517 0.0 13.62546635287517 13.62546635287517 10.0</parameters>
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</space>
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</source>
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<output>
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<summary>false</summary>
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</output>
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<seed>22</seed>
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</settings>
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2
tests/regression_tests/lattice_hex_OX/results_true.dat
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2
tests/regression_tests/lattice_hex_OX/results_true.dat
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k-combined:
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1.355663E+00 2.896562E-02
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171
tests/regression_tests/lattice_hex_OX/test.py
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tests/regression_tests/lattice_hex_OX/test.py
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from tests.testing_harness import PyAPITestHarness
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import openmc
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import numpy as np
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class HexLatticeOXTestHarness(PyAPITestHarness):
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def _build_inputs(self):
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materials = openmc.Materials()
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fuel_mat = openmc.Material(material_id=1,name="UO2")
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fuel_mat.set_density('sum')
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fuel_mat.add_nuclide('U235',0.87370e-03)
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fuel_mat.add_nuclide('U238',1.87440e-02)
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fuel_mat.add_nuclide('O16',3.92350e-02)
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materials.append(fuel_mat)
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coolant = openmc.Material(material_id=2,name="borated H2O")
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coolant.set_density('sum')
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coolant.add_nuclide('H1',0.06694)
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coolant.add_nuclide('O16',0.03347)
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coolant.add_nuclide('B10',6.6262e-6)
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coolant.add_nuclide('B11',2.6839e-5)
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materials.append(coolant)
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absorber = openmc.Material(material_id=3,name="pellet B4C")
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absorber.set_density('sum')
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absorber.add_nuclide('C0',0.01966)
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absorber.add_nuclide('B11',4.7344e-6)
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absorber.add_nuclide('B10',1.9177e-5)
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materials.append(absorber)
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zirc = openmc.Material(material_id=4,name="Zirc4")
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zirc.set_density('sum')
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zirc.add_element('Zr',4.23e-2)
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materials.append(zirc)
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materials.export_to_xml()
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### Geometry ###
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pin_rad = 0.7 # cm
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assembly_pitch = 1.235 # cm
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hexagonal_pitch = 23.6 # cm
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length = 10.0 # cm
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#Fuel pin surfaces
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cylfuelin = openmc.ZCylinder(surface_id=1,r=0.386)
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cylfuelout = openmc.ZCylinder(surface_id=2,r=0.4582)
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#Fuel cells
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infcell = openmc.Cell(cell_id=1)
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infcell.region = -cylfuelin
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infcell.fill = fuel_mat
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clfcell = openmc.Cell(cell_id=2)
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clfcell.region = -cylfuelout & +cylfuelin
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clfcell.fill = zirc
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outfcell = openmc.Cell(cell_id=3)
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outfcell.region = +cylfuelout
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outfcell.fill = coolant
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#Fuel universe
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fuel_ch_univ = openmc.Universe(universe_id=1,name="Fuel channel",cells=[infcell, clfcell,outfcell])
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#Central tube surfaces
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cyltubein = openmc.ZCylinder(surface_id=3,r=0.45)
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cyltubeout = openmc.ZCylinder(surface_id=4,r=0.5177)
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#Central tube cells
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inctcell = openmc.Cell(cell_id=4)
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inctcell.region = -cyltubein
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inctcell.fill = coolant
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clctcell = openmc.Cell(cell_id=5)
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clctcell.region = -cyltubeout & +cyltubein
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clctcell.fill = zirc
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outctcell = openmc.Cell(cell_id=6)
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outctcell.region = +cyltubeout
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outctcell.fill = coolant
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#Central tubel universe
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tube_ch_univ = openmc.Universe(universe_id=2,name="Central tube channel",cells=[inctcell, clctcell,outctcell])
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#Absorber tube surfaces
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cylabsin = openmc.ZCylinder(surface_id=5,r=0.35)
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cylabsout = openmc.ZCylinder(surface_id=6,r=0.41)
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cylabsclin = openmc.ZCylinder(surface_id=7,r=0.545)
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cylabsclout = openmc.ZCylinder(surface_id=8,r=0.6323)
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#Absorber tube cells
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inabscell = openmc.Cell(cell_id=7)
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inabscell.region = -cylabsin
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inabscell.fill = absorber
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clabscell = openmc.Cell(cell_id=8)
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clabscell.region = -cylabsout & +cylabsin
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clabscell.fill = zirc
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interabscell = openmc.Cell(cell_id=9)
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interabscell.region = -cylabsclin & +cylabsout
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interabscell.fill = coolant
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clatcell = openmc.Cell(cell_id=10)
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clatcell.region = -cylabsclout & +cylabsclin
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clatcell.fill = zirc
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outabscell = openmc.Cell(cell_id=11)
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outabscell.region = +cylabsclout
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outabscell.fill = coolant
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#Absorber tube universe
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abs_ch_univ = openmc.Universe(universe_id=3,name="Central tube channel",cells=[inabscell,clabscell,interabscell,clatcell,outabscell])
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#Assembly surfaces
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edge_length = (1./np.sqrt(3.0)) * hexagonal_pitch
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fuel_bottom = openmc.ZPlane(surface_id=9,z0=0.0, boundary_type = 'reflective')
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fuel_top = openmc.ZPlane(surface_id=10,z0=length, boundary_type = 'reflective')
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# a hex surface for the core to go inside of
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hexprism = openmc.model.get_hexagonal_prism(edge_length=edge_length,
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origin=(0.0, 0.0),
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boundary_type = 'reflective',
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orientation='x')
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region = hexprism & +fuel_bottom & -fuel_top
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inf_mat = openmc.Cell(cell_id=12)
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inf_mat.fill = coolant
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inf_mat_univ = openmc.Universe(universe_id=4,cells=[inf_mat,])
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# Fill lattice by channels
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nring = 11
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universes = [] # array of rings
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for ring in range(nring - 1,-1,-1):
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arr = []
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arr.append(fuel_ch_univ)
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for cell in range(ring * 6 - 1):
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arr.append(fuel_ch_univ)
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universes.append(arr)
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universes[-1] = [tube_ch_univ]
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channels = [(7,16),(7,13),(7,10),(7,7),(7,4),(7,1),(5,0),\
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(4,33),(5,25),(4,27),(5,20),(4,21),(5,15),\
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(4,15),(5,10),(4,9),(5,5),(4,3)]
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for i,j in channels:
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universes[i][j] = abs_ch_univ
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lattice = openmc.HexLattice(name="regular fuel assembly")
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lattice.orientation = "ox"
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lattice.center = (0., 0., length/2.0)
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lattice.pitch = (assembly_pitch,length/2.0)
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lattice.universes = 2*[universes]
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lattice.outer = inf_mat_univ
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assembly_cell = openmc.Cell(cell_id=13,name="container assembly cell")
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assembly_cell.region = region
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assembly_cell.fill = lattice
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root_univ = openmc.Universe(universe_id=5,name="root universe", cells=[assembly_cell,])
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geom = openmc.Geometry(root_univ)
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geom.export_to_xml()
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### Settings ###
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settings = openmc.Settings()
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settings.run_mode = 'eigenvalue'
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source = openmc.Source()
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ll = [-edge_length, -edge_length, 0.0]
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ur = [edge_length, edge_length, 10.0]
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source.space = openmc.stats.Box(ll, ur)
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source.strength = 1.0
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settings.source = source
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settings.output = {'summary' : False}
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settings.batches = 10
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settings.inactive = 5
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settings.particles = 1000
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settings.seed = 22
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settings.export_to_xml()
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def test_lattice_hex_ox_surf():
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harness = HexLatticeOXTestHarness('statepoint.10.h5')
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harness.main()
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