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https://github.com/openmc-dev/openmc.git
synced 2026-07-26 21:25:36 -04:00
Added test for making an MGXS lib with nuclides, and then re-running in MG mode. Prior to this we only tested macroscopic MGXS
This commit is contained in:
Adam G Nelson
parent
524b5344eb
commit
2903970129
4 changed files with 338 additions and 0 deletions
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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" name="Fuel" region="-1" universe="0" />
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<cell id="2" material="2" name="Cladding" region="1 -2" universe="0" />
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<cell id="3" material="3" name="Water" region="2 3 -4 5 -6" universe="0" />
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<surface coeffs="0 0 0.39218" id="1" name="Fuel OR" type="z-cylinder" />
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<surface coeffs="0 0 0.4572" id="2" name="Clad OR" type="z-cylinder" />
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<surface boundary="reflective" coeffs="-0.63" id="3" name="left" type="x-plane" />
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<surface boundary="reflective" coeffs="0.63" id="4" name="right" type="x-plane" />
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<surface boundary="reflective" coeffs="-0.63" id="5" name="bottom" type="y-plane" />
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<surface boundary="reflective" coeffs="0.63" id="6" name="top" type="y-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 (2.4%)">
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<density units="g/cm3" value="10.29769" />
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<nuclide ao="4.4843e-06" name="U234" />
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<nuclide ao="0.00055815" name="U235" />
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<nuclide ao="0.022408" name="U238" />
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<nuclide ao="0.045829" name="O16" />
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</material>
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<material id="2" name="Zircaloy">
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<density units="g/cm3" value="6.55" />
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<nuclide ao="0.021827" name="Zr90" />
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<nuclide ao="0.00476" name="Zr91" />
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<nuclide ao="0.0072758" name="Zr92" />
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<nuclide ao="0.0073734" name="Zr94" />
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<nuclide ao="0.0011879" name="Zr96" />
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</material>
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<material id="3" name="Hot borated water">
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<density units="g/cm3" value="0.740582" />
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<nuclide ao="0.049457" name="H1" />
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<nuclide ao="0.024672" name="O16" />
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<nuclide ao="8.0042e-06" name="B10" />
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<nuclide ao="3.2218e-05" name="B11" />
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<sab name="c_H_in_H2O" />
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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>100</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="fission">
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<parameters>-0.63 -0.63 -1 0.63 0.63 1</parameters>
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</space>
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</source>
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</settings>
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<?xml version='1.0' encoding='utf-8'?>
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<tallies>
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<filter id="1" type="material">
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<bins>1</bins>
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</filter>
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<filter id="2" type="energy">
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<bins>0.0 0.625 20000000.0</bins>
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</filter>
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<filter id="7" type="energyout">
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<bins>0.0 0.625 20000000.0</bins>
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</filter>
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<filter id="11" type="legendre">
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<order>3</order>
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</filter>
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<filter id="15" type="material">
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<bins>2</bins>
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</filter>
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<filter id="29" type="material">
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<bins>3</bins>
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</filter>
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<tally id="1">
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<filters>1 2</filters>
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<nuclides>total</nuclides>
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<scores>flux</scores>
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<estimator>tracklength</estimator>
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</tally>
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<tally id="2">
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<filters>1 2</filters>
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<nuclides>U234 U235 U238 O16</nuclides>
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<scores>total</scores>
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<estimator>tracklength</estimator>
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</tally>
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<tally id="3">
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<filters>1 2</filters>
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<nuclides>total</nuclides>
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<scores>flux</scores>
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<estimator>tracklength</estimator>
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</tally>
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<tally id="4">
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<filters>1 2</filters>
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<nuclides>U234 U235 U238 O16</nuclides>
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<scores>absorption</scores>
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<estimator>tracklength</estimator>
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</tally>
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<tally id="5">
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<filters>1 2</filters>
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<nuclides>total</nuclides>
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<scores>flux</scores>
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<estimator>analog</estimator>
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</tally>
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<tally id="6">
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<filters>1 2 7</filters>
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<nuclides>U234 U235 U238 O16</nuclides>
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<scores>nu-fission</scores>
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<estimator>analog</estimator>
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</tally>
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<tally id="7">
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<filters>1 2</filters>
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<nuclides>total</nuclides>
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<scores>flux</scores>
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<estimator>analog</estimator>
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</tally>
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<tally id="8">
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<filters>1 2 7 11</filters>
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<nuclides>U234 U235 U238 O16</nuclides>
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<scores>nu-scatter</scores>
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<estimator>analog</estimator>
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</tally>
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<tally id="9">
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<filters>1 2 7</filters>
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<nuclides>U234 U235 U238 O16</nuclides>
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<scores>nu-scatter</scores>
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<estimator>analog</estimator>
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</tally>
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<tally id="10">
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<filters>1 2 7</filters>
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<nuclides>U234 U235 U238 O16</nuclides>
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<scores>scatter</scores>
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<estimator>analog</estimator>
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</tally>
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<tally id="11">
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<filters>15 2</filters>
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<nuclides>total</nuclides>
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<scores>flux</scores>
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<estimator>tracklength</estimator>
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</tally>
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<tally id="12">
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<filters>15 2</filters>
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<nuclides>Zr90 Zr91 Zr92 Zr94 Zr96</nuclides>
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<scores>total</scores>
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<estimator>tracklength</estimator>
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</tally>
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<tally id="13">
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<filters>15 2</filters>
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<nuclides>total</nuclides>
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<scores>flux</scores>
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<estimator>tracklength</estimator>
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</tally>
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<tally id="14">
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<filters>15 2</filters>
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<nuclides>Zr90 Zr91 Zr92 Zr94 Zr96</nuclides>
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<scores>absorption</scores>
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<estimator>tracklength</estimator>
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</tally>
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<tally id="15">
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<filters>15 2</filters>
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<nuclides>total</nuclides>
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<scores>flux</scores>
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<estimator>analog</estimator>
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</tally>
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<tally id="16">
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<filters>15 2 7</filters>
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<nuclides>Zr90 Zr91 Zr92 Zr94 Zr96</nuclides>
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<scores>nu-fission</scores>
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<estimator>analog</estimator>
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</tally>
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<tally id="17">
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<filters>15 2</filters>
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<nuclides>total</nuclides>
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<scores>flux</scores>
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<estimator>analog</estimator>
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</tally>
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<tally id="18">
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<filters>15 2 7 11</filters>
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<nuclides>Zr90 Zr91 Zr92 Zr94 Zr96</nuclides>
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<scores>nu-scatter</scores>
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<estimator>analog</estimator>
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</tally>
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<tally id="19">
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<filters>15 2 7</filters>
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<nuclides>Zr90 Zr91 Zr92 Zr94 Zr96</nuclides>
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<scores>nu-scatter</scores>
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<estimator>analog</estimator>
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</tally>
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<tally id="20">
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<filters>15 2 7</filters>
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<nuclides>Zr90 Zr91 Zr92 Zr94 Zr96</nuclides>
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<scores>scatter</scores>
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<estimator>analog</estimator>
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</tally>
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<tally id="21">
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<filters>29 2</filters>
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<nuclides>total</nuclides>
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<scores>flux</scores>
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<estimator>tracklength</estimator>
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</tally>
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<tally id="22">
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<filters>29 2</filters>
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<nuclides>H1 O16 B10 B11</nuclides>
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<scores>total</scores>
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<estimator>tracklength</estimator>
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</tally>
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<tally id="23">
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<filters>29 2</filters>
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<nuclides>total</nuclides>
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<scores>flux</scores>
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<estimator>tracklength</estimator>
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</tally>
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<tally id="24">
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<filters>29 2</filters>
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<nuclides>H1 O16 B10 B11</nuclides>
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<scores>absorption</scores>
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<estimator>tracklength</estimator>
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</tally>
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<tally id="25">
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<filters>29 2</filters>
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<nuclides>total</nuclides>
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<scores>flux</scores>
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<estimator>analog</estimator>
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</tally>
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<tally id="26">
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<filters>29 2 7</filters>
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<nuclides>H1 O16 B10 B11</nuclides>
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<scores>nu-fission</scores>
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<estimator>analog</estimator>
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</tally>
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<tally id="27">
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<filters>29 2</filters>
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<nuclides>total</nuclides>
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<scores>flux</scores>
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<estimator>analog</estimator>
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</tally>
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<tally id="28">
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<filters>29 2 7 11</filters>
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<nuclides>H1 O16 B10 B11</nuclides>
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<scores>nu-scatter</scores>
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<estimator>analog</estimator>
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</tally>
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<tally id="29">
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<filters>29 2 7</filters>
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<nuclides>H1 O16 B10 B11</nuclides>
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<scores>nu-scatter</scores>
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<estimator>analog</estimator>
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</tally>
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<tally id="30">
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<filters>29 2 7</filters>
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<nuclides>H1 O16 B10 B11</nuclides>
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<scores>scatter</scores>
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<estimator>analog</estimator>
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</tally>
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</tallies>
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@ -0,0 +1,2 @@
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k-combined:
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5.350442E-01 1.484490E-02
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@ -0,0 +1,85 @@
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import os
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import openmc
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import openmc.mgxs
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from openmc.examples import pwr_pin_cell
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from tests.testing_harness import PyAPITestHarness
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from tests.regression_tests import config
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class MGXSTestHarness(PyAPITestHarness):
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def __init__(self, *args, **kwargs):
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# Generate inputs using parent class routine
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super().__init__(*args, **kwargs)
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# Initialize a two-group structure
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energy_groups = openmc.mgxs.EnergyGroups(group_edges=[0, 0.625, 20.e6])
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# Initialize MGXS Library for a few cross section types
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self.mgxs_lib = openmc.mgxs.Library(self._model.geometry)
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self.mgxs_lib.by_nuclide = True
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self.mgxs_lib.mgxs_types = ['total', 'absorption', 'nu-fission matrix',
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'nu-scatter matrix', 'multiplicity matrix']
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self.mgxs_lib.energy_groups = energy_groups
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self.mgxs_lib.correction = None
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self.mgxs_lib.legendre_order = 3
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self.mgxs_lib.domain_type = 'material'
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self.mgxs_lib.build_library()
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# Initialize a tallies file
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self.mgxs_lib.add_to_tallies_file(self._model.tallies, merge=False)
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def _run_openmc(self):
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# Initial run
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if config['mpi']:
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mpi_args = [config['mpiexec'], '-n', config['mpi_np']]
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openmc.run(openmc_exec=config['exe'], mpi_args=mpi_args)
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else:
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openmc.run(openmc_exec=config['exe'])
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# Build MG Inputs
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# Get data needed to execute Library calculations.
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sp = openmc.StatePoint(self._sp_name)
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self.mgxs_lib.load_from_statepoint(sp)
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self._model.mgxs_file, self._model.materials, \
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self._model.geometry = self.mgxs_lib.create_mg_mode()
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# Modify materials and settings so we can run in MG mode
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self._model.materials.cross_sections = './mgxs.h5'
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self._model.settings.energy_mode = 'multi-group'
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# Write modified input files
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self._model.settings.export_to_xml()
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self._model.geometry.export_to_xml()
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self._model.materials.export_to_xml()
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self._model.mgxs_file.export_to_hdf5()
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# Dont need tallies.xml, so remove the file
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if os.path.exists('tallies.xml'):
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os.remove('tallies.xml')
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# Enforce closing statepoint and summary files so HDF5
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# does not throw an error during the next OpenMC execution
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sp._f.close()
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sp._summary._f.close()
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# Re-run MG mode.
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if config['mpi']:
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mpi_args = [config['mpiexec'], '-n', config['mpi_np']]
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openmc.run(openmc_exec=config['exe'], mpi_args=mpi_args)
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else:
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openmc.run(openmc_exec=config['exe'])
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def _cleanup(self):
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super()._cleanup()
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f = 'mgxs.h5'
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if os.path.exists(f):
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os.remove(f)
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def test_mgxs_library_ce_to_mg():
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# Set the input set to use the pincell model
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model = pwr_pin_cell()
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harness = MGXSTestHarness('statepoint.10.h5', model)
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harness.main()
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