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Merge pull request #2432 from guyshtot/library_nuclides_specification
Library nuclides specification
This commit is contained in:
Paul Romano
GitHub
commit
99c6514609
5 changed files with 322 additions and 2 deletions
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@ -72,6 +72,11 @@ class Library:
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Number of equi-width polar angle bins for angle discretization
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num_azimuthal : Integral
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Number of equi-width azimuthal angle bins for angle discretization
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nuclides : Iterable of str or 'sum'
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The optional user-specified nuclides for which to compute cross
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sections (e.g., 'U238', 'O16'). If by_nuclide is True but nuclides
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are not specified by the user, all nuclides in the domain
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are included.
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estimator : str or None
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The tally estimator used to compute multi-group cross sections.
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If None, the default for each MGXS type is used.
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@ -107,6 +112,7 @@ class Library:
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self._energy_groups = None
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self._num_polar = 1
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self._num_azimuthal = 1
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self._nuclides = None
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self._num_delayed_groups = 0
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self._correction = 'P0'
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self._scatter_format = 'legendre'
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@ -145,6 +151,7 @@ class Library:
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clone._energy_groups = copy.deepcopy(self.energy_groups, memo)
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clone._num_polar = self.num_polar
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clone._num_azimuthal = self.num_azimuthal
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clone._nuclides = self._nuclides
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clone._num_delayed_groups = self.num_delayed_groups
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clone._tally_trigger = copy.deepcopy(self.tally_trigger, memo)
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clone._all_mgxs = copy.deepcopy(self.all_mgxs)
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@ -205,6 +212,10 @@ class Library:
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else:
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return self._domains
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@property
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def nuclides(self):
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return self._nuclides
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@property
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def energy_groups(self):
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return self._energy_groups
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@ -275,6 +286,11 @@ class Library:
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cv.check_type('name', name, str)
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self._name = name
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@nuclides.setter
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def nuclides(self, nuclides):
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cv.check_iterable_type('nuclides', nuclides, str)
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self._nuclides = nuclides
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@mgxs_types.setter
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def mgxs_types(self, mgxs_types):
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all_mgxs_types = openmc.mgxs.MGXS_TYPES + openmc.mgxs.MDGXS_TYPES + \
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@ -524,6 +540,20 @@ class Library:
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mgxs.legendre_order = self.legendre_order
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mgxs.histogram_bins = self.histogram_bins
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if self.by_nuclide:
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try:
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domain_nuclides = domain.get_nuclides()
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except AttributeError:
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domain_nuclides = None
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if self.nuclides:
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if domain_nuclides:
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mgxs.nuclides = [
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nuclide for nuclide in self.nuclides
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if nuclide in domain_nuclides
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] + ["total"]
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else:
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mgxs.nuclides = self.nuclides
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self.all_mgxs[domain.id][mgxs_type] = mgxs
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def add_to_tallies_file(self, tallies_file, merge=True):
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@ -590,7 +620,7 @@ class Library:
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self._sp_filename = statepoint._f.filename
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self._geometry = statepoint.summary.geometry
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self._nuclides = statepoint.summary.nuclides
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self._atomic_weight_ratios = statepoint.summary.nuclides
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if statepoint.run_mode == 'eigenvalue':
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self._keff = statepoint.keff.n
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@ -1005,7 +1035,7 @@ class Library:
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xsdata.num_azimuthal = self.num_azimuthal
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if nuclide != 'total':
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xsdata.atomic_weight_ratio = self._nuclides[nuclide]
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xsdata.atomic_weight_ratio = self._atomic_weight_ratios[nuclide]
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if subdomain is None:
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subdomain = 'all'
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@ -0,0 +1,218 @@
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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="383" type="material">
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<bins>1 2 3</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="1" type="material">
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<bins>1</bins>
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</filter>
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<filter id="5" 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="6" type="legendre">
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<order>1</order>
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</filter>
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<filter id="30" type="legendre">
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<order>3</order>
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</filter>
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<filter id="54" type="energy">
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<bins>0.0 20000000.0</bins>
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</filter>
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<filter id="106" type="material">
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<bins>2</bins>
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</filter>
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<filter id="251" type="material">
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<bins>3</bins>
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</filter>
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<tally id="528">
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<filters>383 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="167">
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<filters>1 2</filters>
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<nuclides>U235 total</nuclides>
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<scores>total absorption (n,2n) (n,3n) (n,4n) fission nu-fission kappa-fission scatter inverse-velocity prompt-nu-fission (n,elastic) (n,level) (n,na) (n,nc) (n,gamma) (n,a) (n,Xa) heating damage-energy (n,n1) (n,a0)</scores>
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<estimator>tracklength</estimator>
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</tally>
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<tally id="540">
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<filters>383 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="119">
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<filters>1 5 6</filters>
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<nuclides>U235 total</nuclides>
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<scores>scatter nu-scatter</scores>
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<estimator>analog</estimator>
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</tally>
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<tally id="54">
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<filters>1 2</filters>
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<nuclides>U235 total</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="84">
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<filters>1 2 5 30</filters>
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<nuclides>U235 total</nuclides>
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<scores>scatter nu-scatter</scores>
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<estimator>analog</estimator>
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</tally>
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<tally id="179">
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<filters>1 2 5</filters>
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<nuclides>U235 total</nuclides>
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<scores>nu-scatter scatter nu-fission prompt-nu-fission (n,nc) (n,n1) (n,2n)</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>1 54</filters>
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<nuclides>U235 total</nuclides>
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<scores>nu-fission prompt-nu-fission</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>1 5</filters>
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<nuclides>U235 total</nuclides>
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<scores>nu-fission prompt-nu-fission</scores>
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<estimator>analog</estimator>
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</tally>
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<tally id="348">
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<filters>106 2</filters>
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<nuclides>Zr90 total</nuclides>
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<scores>total absorption (n,2n) (n,3n) (n,4n) fission nu-fission kappa-fission scatter inverse-velocity prompt-nu-fission (n,elastic) (n,level) (n,na) (n,nc) (n,gamma) (n,a) (n,Xa) heating damage-energy (n,n1) (n,a0)</scores>
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<estimator>tracklength</estimator>
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</tally>
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<tally id="300">
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<filters>106 5 6</filters>
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<nuclides>Zr90 total</nuclides>
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<scores>scatter nu-scatter</scores>
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<estimator>analog</estimator>
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</tally>
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<tally id="235">
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<filters>106 2</filters>
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<nuclides>Zr90 total</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="265">
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<filters>106 2 5 30</filters>
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<nuclides>Zr90 total</nuclides>
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<scores>scatter nu-scatter</scores>
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<estimator>analog</estimator>
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</tally>
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<tally id="360">
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<filters>106 2 5</filters>
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<nuclides>Zr90 total</nuclides>
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<scores>nu-scatter scatter nu-fission prompt-nu-fission (n,nc) (n,n1) (n,2n)</scores>
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<estimator>analog</estimator>
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</tally>
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<tally id="271">
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<filters>106 54</filters>
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<nuclides>Zr90 total</nuclides>
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<scores>nu-fission prompt-nu-fission</scores>
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<estimator>analog</estimator>
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</tally>
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<tally id="272">
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<filters>106 5</filters>
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<nuclides>Zr90 total</nuclides>
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<scores>nu-fission prompt-nu-fission</scores>
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<estimator>analog</estimator>
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</tally>
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<tally id="529">
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<filters>251 2</filters>
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<nuclides>H1 total</nuclides>
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<scores>total absorption (n,2n) (n,3n) (n,4n) fission nu-fission kappa-fission scatter inverse-velocity prompt-nu-fission (n,elastic) (n,level) (n,na) (n,nc) (n,gamma) (n,a) (n,Xa) heating damage-energy (n,n1) (n,a0)</scores>
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<estimator>tracklength</estimator>
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</tally>
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<tally id="481">
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<filters>251 5 6</filters>
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<nuclides>H1 total</nuclides>
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<scores>scatter nu-scatter</scores>
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<estimator>analog</estimator>
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</tally>
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<tally id="416">
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<filters>251 2</filters>
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<nuclides>H1 total</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="446">
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<filters>251 2 5 30</filters>
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<nuclides>H1 total</nuclides>
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<scores>scatter nu-scatter</scores>
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<estimator>analog</estimator>
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</tally>
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<tally id="541">
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<filters>251 2 5</filters>
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<nuclides>H1 total</nuclides>
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<scores>nu-scatter scatter nu-fission prompt-nu-fission (n,nc) (n,n1) (n,2n)</scores>
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<estimator>analog</estimator>
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</tally>
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<tally id="452">
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<filters>251 54</filters>
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<nuclides>H1 total</nuclides>
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<scores>nu-fission prompt-nu-fission</scores>
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<estimator>analog</estimator>
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</tally>
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<tally id="453">
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<filters>251 5</filters>
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<nuclides>H1 total</nuclides>
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<scores>nu-fission prompt-nu-fission</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 @@
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3e86542d1166b8a0bcc61742b88c9e931d63733477f3274b32b4da64d8f05413fa6330d5615f44f24735c2c2f77d3071f3dce38faba284c321ceab81c1064480
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@ -0,0 +1,71 @@
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import hashlib
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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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class MGXSTestHarness(PyAPITestHarness):
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def __init__(self, *args, **kwargs):
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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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# Test relevant MGXS types
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relevant_MGXS_TYPES = [item for item in openmc.mgxs.MGXS_TYPES
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if item != 'current']
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# Add in a subset of openmc.mgxs.ARBITRARY_VECTOR_TYPES and
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# openmc.mgxs.ARBITRARY_MATRIX_TYPES so we can see the code works,
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# but not use too much resources
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relevant_MGXS_TYPES += [
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"(n,elastic)", "(n,level)", "(n,2n)", "(n,na)", "(n,nc)",
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"(n,gamma)", "(n,a)", "(n,Xa)", "heating", "damage-energy",
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"(n,n1)", "(n,a0)", "(n,nc) matrix", "(n,n1) matrix",
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"(n,2n) matrix"]
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self.mgxs_lib.mgxs_types = tuple(relevant_MGXS_TYPES)
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self.mgxs_lib.energy_groups = energy_groups
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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.nuclides = ['U235', 'Zr90', 'H1']
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self.mgxs_lib.build_library()
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# Add tallies
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self.mgxs_lib.add_to_tallies_file(self._model.tallies, merge=True)
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def _get_results(self, hash_output=True):
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"""Digest info in the statepoint and return as a string."""
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# Read the statepoint file.
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sp = openmc.StatePoint(self._sp_name)
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# Load the MGXS library from the statepoint
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self.mgxs_lib.load_from_statepoint(sp)
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# Build a string from Pandas Dataframe for each MGXS
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outstr = ''
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for domain in self.mgxs_lib.domains:
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for mgxs_type in self.mgxs_lib.mgxs_types:
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mgxs = self.mgxs_lib.get_mgxs(domain, mgxs_type)
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df = mgxs.get_pandas_dataframe()
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outstr += df.to_string() + '\n'
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# Hash the results if necessary
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if hash_output:
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sha512 = hashlib.sha512()
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sha512.update(outstr.encode('utf-8'))
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outstr = sha512.hexdigest()
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return outstr
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def test_mgxs_library_specific_nuclides():
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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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