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Fix undefined variable in openmc.mgxs.library (scatt_mgxs) (#3943)
Co-authored-by: Paul Romano <paul.k.romano@gmail.com>
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hugo-barthod
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4 changed files with 244 additions and 5 deletions
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@ -1290,6 +1290,14 @@ class Library:
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'are ignored since multiplicity or nu-scatter matrices '\
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'were not tallied for ' + xsdata_name
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warn(msg, RuntimeWarning)
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if 'scatter matrix' in self.mgxs_types:
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scatt_mgxs = self.get_mgxs(domain, 'scatter matrix')
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elif 'consistent scatter matrix' in self.mgxs_types:
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scatt_mgxs = self.get_mgxs(domain, 'consistent scatter matrix')
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else:
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raise ValueError(f'No scatter matrix found for {xsdata_name}.')
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xsdata.set_scatter_matrix_mgxs(scatt_mgxs, temperature=temperature,
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xs_type=xs_type,
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nuclide=[nuclide],
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@ -0,0 +1,217 @@
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<?xml version='1.0' encoding='utf-8'?>
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<model>
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<materials>
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<material id="7" name="UO2 (2.4%)" depletable="true">
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<density value="10.29769" units="g/cm3"/>
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<nuclide name="U234" ao="4.4843e-06"/>
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<nuclide name="U235" ao="0.00055815"/>
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<nuclide name="U238" ao="0.022408"/>
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<nuclide name="O16" ao="0.045829"/>
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</material>
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<material id="8" name="Zircaloy">
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<density value="6.55" units="g/cm3"/>
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<nuclide name="Zr90" ao="0.021827"/>
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<nuclide name="Zr91" ao="0.00476"/>
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<nuclide name="Zr92" ao="0.0072758"/>
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<nuclide name="Zr94" ao="0.0073734"/>
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<nuclide name="Zr96" ao="0.0011879"/>
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</material>
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<material id="9" name="Hot borated water">
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<density value="0.740582" units="g/cm3"/>
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<nuclide name="H1" ao="0.049457"/>
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<nuclide name="O16" ao="0.024672"/>
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<nuclide name="B10" ao="8.0042e-06"/>
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<nuclide name="B11" ao="3.2218e-05"/>
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<sab name="c_H_in_H2O"/>
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</material>
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</materials>
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<geometry>
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<cell id="4" name="Fuel" material="7" region="-7" universe="0"/>
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<cell id="5" name="Cladding" material="8" region="7 -8" universe="0"/>
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<cell id="6" name="Water" material="9" region="8 9 -10 11 -12" universe="0"/>
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<surface id="7" name="Fuel OR" type="z-cylinder" coeffs="0 0 0.39218"/>
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<surface id="8" name="Clad OR" type="z-cylinder" coeffs="0 0 0.4572"/>
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<surface id="9" name="left" type="x-plane" boundary="reflective" coeffs="-0.63"/>
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<surface id="10" name="right" type="x-plane" boundary="reflective" coeffs="0.63"/>
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<surface id="11" name="bottom" type="y-plane" boundary="reflective" coeffs="-0.63"/>
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<surface id="12" name="top" type="y-plane" boundary="reflective" coeffs="0.63"/>
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</geometry>
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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 type="independent" strength="1.0" particle="neutron">
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<space type="box">
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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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<constraints>
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<fissionable>true</fissionable>
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</constraints>
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</source>
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</settings>
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<tallies>
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<filter id="43" type="material">
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<bins>7</bins>
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</filter>
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<filter id="44" 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="49" 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="53" type="legendre">
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<order>3</order>
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</filter>
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<filter id="54" type="material">
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<bins>8</bins>
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</filter>
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<filter id="65" type="material">
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<bins>9</bins>
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</filter>
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<tally id="76">
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<filters>43 44</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="77">
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<filters>43 44</filters>
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<nuclides>total</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="78">
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<filters>43 44</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="79">
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<filters>43 44</filters>
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<nuclides>total</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="80">
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<filters>43 44</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="81">
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<filters>43 44 49</filters>
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<nuclides>total</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="82">
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<filters>43 44</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="83">
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<filters>43 44 49 53</filters>
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<nuclides>total</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="84">
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<filters>54 44</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="85">
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<filters>54 44</filters>
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<nuclides>total</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="86">
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<filters>54 44</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="87">
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<filters>54 44</filters>
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<nuclides>total</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="88">
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<filters>54 44</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="89">
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<filters>54 44 49</filters>
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<nuclides>total</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="90">
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<filters>54 44</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="91">
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<filters>54 44 49 53</filters>
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<nuclides>total</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="92">
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<filters>65 44</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="93">
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<filters>65 44</filters>
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<nuclides>total</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="94">
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<filters>65 44</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="95">
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<filters>65 44</filters>
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<nuclides>total</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="96">
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<filters>65 44</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="97">
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<filters>65 44 49</filters>
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<nuclides>total</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="98">
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<filters>65 44</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="99">
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<filters>65 44 49 53</filters>
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<nuclides>total</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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</model>
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@ -9,7 +9,7 @@ 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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def __init__(self, *args, scatter_mgxs_type=None, **kwargs):
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# Generate inputs using parent class routine
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super().__init__(*args, **kwargs)
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@ -19,8 +19,8 @@ class MGXSTestHarness(PyAPITestHarness):
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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 = False
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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.mgxs_types = ['total', 'absorption', 'nu-fission matrix']
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self.mgxs_lib.mgxs_types += scatter_mgxs_type
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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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@ -69,9 +69,23 @@ class MGXSTestHarness(PyAPITestHarness):
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os.remove(f)
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def test_mgxs_library_ce_to_mg():
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def test_mgxs_library_ce_to_mg_multiplicity_matrix():
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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 = MGXSTestHarness(
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'statepoint.10.h5', model,
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inputs_true='inputs_true_multiplicity_matrix.dat',
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scatter_mgxs_type=['nu-scatter matrix', 'multiplicity matrix']
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)
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harness.main()
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def test_mgxs_library_ce_to_mg_scatter_matrix():
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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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inputs_true='inputs_true_scatter_matrix.dat',
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scatter_mgxs_type=['scatter matrix'])
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harness.main()
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