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Merge pull request #1705 from nelsonag/mt_mgxs
MGXS API Changes for Arbitrary Reaction Channels and for Accurate Chi Computation
This commit is contained in:
Paul Romano
GitHub
commit
7a505613d1
12 changed files with 2838 additions and 1002 deletions
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@ -30,6 +30,7 @@ Multi-group Cross Sections
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:template: myclassinherit.rst
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openmc.mgxs.MGXS
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openmc.mgxs.MatrixMGXS
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openmc.mgxs.AbsorptionXS
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openmc.mgxs.CaptureXS
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openmc.mgxs.Chi
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@ -45,6 +46,9 @@ Multi-group Cross Sections
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openmc.mgxs.ScatterProbabilityMatrix
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openmc.mgxs.TotalXS
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openmc.mgxs.TransportXS
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openmc.mgxs.ArbitraryXS
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openmc.mgxs.ArbitraryMatrixXS
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openmc.mgxs.MeshSurfaceMGXS
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Multi-delayed-group Cross Sections
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----------------------------------
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@ -55,6 +59,7 @@ Multi-delayed-group Cross Sections
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:template: myclassinherit.rst
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openmc.mgxs.MDGXS
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openmc.mgxs.MatrixMDGXS
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openmc.mgxs.ChiDelayed
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openmc.mgxs.DelayedNuFissionXS
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openmc.mgxs.DelayedNuFissionMatrixXS
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@ -199,7 +199,7 @@ enum ReactionType {
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N_3N3HE = 177,
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N_4N3HE = 178,
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N_3N2P = 179,
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N_3N3A = 180,
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N_3N2A = 180,
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N_3NPA = 181,
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N_DT = 182,
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N_NPD = 183,
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@ -56,7 +56,7 @@ REACTION_NAME = {1: '(n,total)', 2: '(n,elastic)', 4: '(n,level)',
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301: 'heating', 444: 'damage-energy',
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649: '(n,pc)', 699: '(n,dc)', 749: '(n,tc)', 799: '(n,3Hec)',
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849: '(n,ac)', 891: '(n,2nc)', 901: 'heating-local'}
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REACTION_NAME.update({i: '(n,n{})'.format(i - 50) for i in range(50, 91)})
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REACTION_NAME.update({i: '(n,n{})'.format(i - 50) for i in range(51, 91)})
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REACTION_NAME.update({i: '(n,p{})'.format(i - 600) for i in range(600, 649)})
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REACTION_NAME.update({i: '(n,d{})'.format(i - 650) for i in range(650, 699)})
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REACTION_NAME.update({i: '(n,t{})'.format(i - 700) for i in range(700, 749)})
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@ -277,7 +277,9 @@ class Library:
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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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all_mgxs_types = openmc.mgxs.MGXS_TYPES + openmc.mgxs.MDGXS_TYPES + \
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openmc.mgxs.ARBITRARY_VECTOR_TYPES + \
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openmc.mgxs.ARBITRARY_MATRIX_TYPES
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if mgxs_types == 'all':
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self._mgxs_types = all_mgxs_types
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else:
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@ -612,8 +614,10 @@ class Library:
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----------
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domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.RegularMesh or Integral
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The material, cell, or universe object of interest (or its ID)
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mgxs_type : {'total', 'transport', 'nu-transport', 'absorption', 'capture', 'fission', 'nu-fission', 'kappa-fission', 'scatter', 'nu-scatter', 'scatter matrix', 'nu-scatter matrix', 'multiplicity matrix', 'nu-fission matrix', chi', 'chi-prompt', 'inverse-velocity', 'prompt-nu-fission', 'prompt-nu-fission matrix', 'delayed-nu-fission', 'delayed-nu-fission matrix', 'chi-delayed', 'beta'}
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The type of multi-group cross section object to return
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mgxs_type : str
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The type of multi-group cross section object to return; allowable
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values are those MGXS to the Library and present in the
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mgxs_types attribute.
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Returns
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-------
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@ -912,7 +916,7 @@ class Library:
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return pickle.load(f)
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def get_xsdata(self, domain, xsdata_name, nuclide='total', xs_type='macro',
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subdomain=None):
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subdomain=None, apply_domain_chi=False):
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"""Generates an openmc.XSdata object describing a multi-group cross section
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dataset for writing to an openmc.MGXSLibrary object.
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@ -939,6 +943,15 @@ class Library:
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mesh cell of interest in the openmc.RegularMesh object. Note:
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this parameter currently only supports subdomains within a mesh,
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and not the subdomains of a distribcell.
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apply_domain_chi : bool
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This parameter sets whether (True) or not (False) the
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domain-averaged values of chi, chi-prompt, and chi-delayed are to
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be applied to each of the nuclide-dependent fission energy spectra
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of a domain. In effect, if this is True, then every nuclide in the
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domain receives the same flux-weighted Chi. This is useful for
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downstream multigroup solvers that precompute a material-specific
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chi before the transport solve provides group-wise fluxes. Defaults
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to False.
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Returns
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-------
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@ -1046,18 +1059,30 @@ class Library:
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if 'chi' in self.mgxs_types:
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mymgxs = self.get_mgxs(domain, 'chi')
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xsdata.set_chi_mgxs(mymgxs, xs_type=xs_type, nuclide=[nuclide],
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if apply_domain_chi and nuclide != "total":
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nuc = "sum"
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else:
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nuc = nuclide
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xsdata.set_chi_mgxs(mymgxs, xs_type=xs_type, nuclide=[nuc],
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subdomain=subdomain)
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if 'chi-prompt' in self.mgxs_types:
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mymgxs = self.get_mgxs(domain, 'chi-prompt')
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if apply_domain_chi and nuclide != "total":
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nuc = "sum"
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else:
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nuc = nuclide
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xsdata.set_chi_prompt_mgxs(mymgxs, xs_type=xs_type,
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nuclide=[nuclide], subdomain=subdomain)
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nuclide=[nuc], subdomain=subdomain)
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if 'chi-delayed' in self.mgxs_types:
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mymgxs = self.get_mgxs(domain, 'chi-delayed')
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if apply_domain_chi and nuclide != "total":
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nuc = "sum"
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else:
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nuc = nuclide
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xsdata.set_chi_delayed_mgxs(mymgxs, xs_type=xs_type,
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nuclide=[nuclide], subdomain=subdomain)
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nuclide=[nuc], subdomain=subdomain)
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if 'nu-fission' in self.mgxs_types:
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mymgxs = self.get_mgxs(domain, 'nu-fission')
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@ -1196,7 +1221,8 @@ class Library:
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return xsdata
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def create_mg_library(self, xs_type='macro', xsdata_names=None):
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def create_mg_library(self, xs_type='macro', xsdata_names=None,
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apply_domain_chi=False):
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"""Creates an openmc.MGXSLibrary object to contain the MGXS data for the
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Multi-Group mode of OpenMC.
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@ -1213,6 +1239,15 @@ class Library:
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xsdata_names : Iterable of str
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List of names to apply to the "xsdata" entries in the
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resultant mgxs data file. Defaults to 'set1', 'set2', ...
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apply_domain_chi : bool
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This parameter sets whether (True) or not (False) the
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domain-averaged values of chi, chi-prompt, and chi-delayed are to
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be applied to each of the nuclide-dependent fission energy spectra
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of a domain. In effect, if this is True, then every nuclide in the
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domain receives the same flux-weighted Chi. This is useful for
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downstream multigroup solvers that precompute a material-specific
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chi before the transport solve provides group-wise fluxes. Defaults
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to False.
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Returns
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-------
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@ -1284,13 +1319,15 @@ class Library:
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xsdata_name = xsdata_names[i]
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xsdata = self.get_xsdata(domain, xsdata_name,
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nuclide=nuclide, xs_type=xs_type)
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nuclide=nuclide, xs_type=xs_type,
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apply_domain_chi=apply_domain_chi)
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mgxs_file.add_xsdata(xsdata)
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return mgxs_file
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def create_mg_mode(self, xsdata_names=None, bc=['reflective'] * 6):
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def create_mg_mode(self, xsdata_names=None, bc=['reflective'] * 6,
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apply_domain_chi=False):
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"""Creates an openmc.MGXSLibrary object to contain the MGXS data for the
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Multi-Group mode of OpenMC as well as the associated openmc.Materials
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and openmc.Geometry objects.
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@ -1314,6 +1351,15 @@ class Library:
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(if applying to a 3D mesh) provided in the following order:
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[x min, x max, y min, y max, z min, z max]. 2-D cells do not
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contain the z min and z max entries.
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apply_domain_chi : bool
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This parameter sets whether (True) or not (False) the
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domain-averaged values of chi, chi-prompt, and chi-delayed are to
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be applied to each of the nuclide-dependent fission energy spectra
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of a domain. In effect, if this is True, then every nuclide in the
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domain receives the same flux-weighted Chi. This is useful for
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downstream multigroup solvers that precompute a material-specific
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chi before the transport solve provides group-wise fluxes. Defaults
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to False.
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Returns
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-------
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@ -1354,7 +1400,8 @@ class Library:
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cv.check_length("domains", self.domains, 1, 1)
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# Get the MGXS File Data
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mgxs_file = self.create_mg_library('macro', xsdata_names)
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mgxs_file = self.create_mg_library('macro', xsdata_names,
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apply_domain_chi=apply_domain_chi)
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# Now move on the creating the geometry and assigning materials
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if self.domain_type == 'mesh':
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@ -1415,7 +1462,7 @@ class Library:
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if not isinstance(cell.fill, openmc.Material):
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warn('If the library domain includes a lattice or universe cell '
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'in conjunction with a consituent cell of that lattice/universe, '
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'the multi-group simulation will fail')
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'the multi-group simulation will fail')
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if cell.id == domain.id:
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cell.fill = material
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@ -8,6 +8,7 @@ import h5py
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import numpy as np
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import openmc
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from openmc.data import REACTION_MT, REACTION_NAME, FISSION_MTS
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import openmc.checkvalue as cv
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from ..tallies import ESTIMATOR_TYPES
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from . import EnergyGroups
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@ -42,6 +43,22 @@ MGXS_TYPES = (
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'nu-diffusion-coefficient'
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)
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# Some scores from REACTION_MT are not supported, or are simply overkill to
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# support and test (like inelastic levels), remoev those from consideration
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_BAD_SCORES = ["(n,misc)", "(n,absorption)", "(n,total)", "fission"]
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_BAD_SCORES += [REACTION_NAME[mt] for mt in FISSION_MTS]
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ARBITRARY_VECTOR_TYPES = tuple(k for k in REACTION_MT.keys()
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if k not in _BAD_SCORES)
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ARBITRARY_MATRIX_TYPES = []
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for rxn in ARBITRARY_VECTOR_TYPES:
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# Preclude the fission channels from being treated as a matrix
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if rxn not in [REACTION_NAME[mt] for mt in FISSION_MTS]:
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split_rxn = rxn.strip("()").split(",")
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if len(split_rxn) > 1 and "n" in split_rxn[1]:
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# Then there is a neutron product, so it can also be a matrix
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ARBITRARY_MATRIX_TYPES.append(rxn + " matrix")
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ARBITRARY_MATRIX_TYPES = tuple(ARBITRARY_MATRIX_TYPES)
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# Supported domain types
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DOMAIN_TYPES = (
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'cell',
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@ -698,8 +715,13 @@ class MGXS:
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Parameters
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----------
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mgxs_type : {'total', 'transport', 'nu-transport', 'absorption', 'capture', 'fission', 'nu-fission', 'kappa-fission', 'scatter', 'nu-scatter', 'scatter matrix', 'nu-scatter matrix', 'multiplicity matrix', 'nu-fission matrix', 'chi', 'chi-prompt', 'inverse-velocity', 'prompt-nu-fission', 'prompt-nu-fission matrix', 'current', 'diffusion-coefficient', 'nu-diffusion-coefficient'}
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The type of multi-group cross section object to return
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mgxs_type : str or Integral
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The type of multi-group cross section object to return; valid
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values are members of MGXS_TYPES, or the reaction types that are
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the keys of REACTION_MT. Note that if a reaction type from
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REACTION_MT is used, it can be appended with ' matrix' to obtain
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a multigroup matrix (from incoming to outgoing energy groups) for
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reactions with a neutron in an outgoing channel.
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domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.RegularMesh
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The domain for spatial homogenization
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domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
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@ -727,7 +749,9 @@ class MGXS:
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"""
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cv.check_value('mgxs_type', mgxs_type, MGXS_TYPES)
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cv.check_value(
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"mgxs_type", mgxs_type,
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MGXS_TYPES + ARBITRARY_VECTOR_TYPES + ARBITRARY_MATRIX_TYPES)
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if mgxs_type == 'total':
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mgxs = TotalXS(domain, domain_type, energy_groups)
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@ -782,6 +806,13 @@ class MGXS:
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mgxs = DiffusionCoefficient(domain, domain_type, energy_groups)
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elif mgxs_type == 'nu-diffusion-coefficient':
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mgxs = DiffusionCoefficient(domain, domain_type, energy_groups, nu=True)
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elif mgxs_type in ARBITRARY_VECTOR_TYPES:
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# Then it is a reaction not covered by the above that is
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# supported by the ArbitraryXS Class
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mgxs = ArbitraryXS(mgxs_type, domain, domain_type, energy_groups)
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elif mgxs_type in ARBITRARY_MATRIX_TYPES:
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mgxs = ArbitraryMatrixXS(mgxs_type, domain, domain_type,
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energy_groups)
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mgxs.by_nuclide = by_nuclide
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mgxs.name = name
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@ -1139,6 +1170,119 @@ class MGXS:
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return xs
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def get_flux(self, groups='all', subdomains='all',
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order_groups='increasing', value='mean',
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squeeze=True, **kwargs):
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r"""Returns an array of the fluxes used to weight the MGXS.
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This method constructs a 2D NumPy array for the requested
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weighting flux for one or more subdomains (1st dimension), and
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energy groups (2nd dimension).
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Parameters
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----------
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groups : Iterable of Integral or 'all'
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Energy groups of interest. Defaults to 'all'.
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subdomains : Iterable of Integral or 'all'
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Subdomain IDs of interest. Defaults to 'all'.
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order_groups: {'increasing', 'decreasing'}
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Return the cross section indexed according to increasing or
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decreasing energy groups (decreasing or increasing energies).
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Defaults to 'increasing'.
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value : {'mean', 'std_dev', 'rel_err'}
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A string for the type of value to return. Defaults to 'mean'.
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squeeze : bool
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A boolean representing whether to eliminate the extra dimensions
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of the multi-dimensional array to be returned. Defaults to True.
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Returns
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-------
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numpy.ndarray
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A NumPy array of the flux indexed in the order
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each group and subdomain is listed in the parameters.
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Raises
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------
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ValueError
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When this method is called before the data is available from tally
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data, or, when this is used on an MGXS type without a flux score.
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"""
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cv.check_value('value', value, ['mean', 'std_dev', 'rel_err'])
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filters = []
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filter_bins = []
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# Construct a collection of the domain filter bins
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if not isinstance(subdomains, str):
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cv.check_iterable_type('subdomains', subdomains, Integral,
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max_depth=3)
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filters.append(_DOMAIN_TO_FILTER[self.domain_type])
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subdomain_bins = []
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for subdomain in subdomains:
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subdomain_bins.append(subdomain)
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filter_bins.append(tuple(subdomain_bins))
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# Construct list of energy group bounds tuples for all requested groups
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if not isinstance(groups, str):
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cv.check_iterable_type('groups', groups, Integral)
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filters.append(openmc.EnergyFilter)
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energy_bins = []
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for group in groups:
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energy_bins.append(
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(self.energy_groups.get_group_bounds(group),))
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filter_bins.append(tuple(energy_bins))
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# Determine which flux to obtain
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# Step through in order of usefulness
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for key in ['flux', 'flux (tracklength)', 'flux (analog)']:
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if key in self.tally_keys:
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tally = self.tallies[key]
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break
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else:
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msg = "MGXS of Type {} do not have an explicit weighting flux!"
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raise ValueError(msg.format(self.__name__))
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flux = tally.get_values(filters=filters, filter_bins=filter_bins,
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nuclides=['total'], value=value)
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# Eliminate the trivial score dimension
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flux = np.squeeze(flux, axis=len(flux.shape) - 1)
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# Eliminate the trivial nuclide dimension
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flux = np.squeeze(flux, axis=len(flux.shape) - 1)
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flux = np.nan_to_num(flux)
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if groups == 'all':
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num_groups = self.num_groups
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else:
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num_groups = len(groups)
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# Reshape tally data array with separate axes for domain and energy
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# Accomodate the polar and azimuthal bins if needed
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num_subdomains = int(flux.shape[0] / (num_groups * self.num_polar *
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self.num_azimuthal))
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if self.num_polar > 1 or self.num_azimuthal > 1:
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new_shape = (self.num_polar, self.num_azimuthal, num_subdomains,
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num_groups)
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else:
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new_shape = (num_subdomains, num_groups)
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new_shape += flux.shape[1:]
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flux = np.reshape(flux, new_shape)
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# Reverse data if user requested increasing energy groups since
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# tally data is stored in order of increasing energies
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if order_groups == 'increasing':
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flux = flux[..., ::-1]
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if squeeze:
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# We want to squeeze out everything but the polar, azimuthal,
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# and energy group data.
|
||||
flux = self._squeeze_xs(flux)
|
||||
|
||||
return flux
|
||||
|
||||
def get_condensed_xs(self, coarse_groups):
|
||||
"""Construct an energy-condensed version of this cross section.
|
||||
|
||||
|
|
@ -3850,6 +3994,265 @@ class ScatterXS(MGXS):
|
|||
self._valid_estimators = ['analog']
|
||||
|
||||
|
||||
class ArbitraryXS(MGXS):
|
||||
r"""A multi-group cross section for an arbitrary reaction type.
|
||||
|
||||
This class can be used for both OpenMC input generation and tally data
|
||||
post-processing to compute spatially-homogenized and energy-integrated
|
||||
multi-group total cross sections for multi-group neutronics calculations.
|
||||
At a minimum, one needs to set the :attr:`ArbitraryXS.energy_groups` and
|
||||
:attr:`ArbitraryXS.domain` properties. Tallies for the flux and appropriate
|
||||
reaction rates over the specified domain are generated automatically via the
|
||||
:attr:`ArbitraryXS.tallies` property, which can then be appended to a
|
||||
:class:`openmc.Tallies` instance.
|
||||
|
||||
For post-processing, the :meth:`MGXS.load_from_statepoint` will pull in the
|
||||
necessary data to compute multi-group cross sections from a
|
||||
:class:`openmc.StatePoint` instance. The derived multi-group cross section
|
||||
can then be obtained from the :attr:`ArbitraryXS.xs_tally` property.
|
||||
|
||||
For a spatial domain :math:`V` and energy group :math:`[E_g,E_{g-1}]`, the
|
||||
requested cross section is calculated as:
|
||||
|
||||
.. math::
|
||||
|
||||
\frac{\int_{r \in V} dr \int_{4\pi} d\Omega \int_{E_g}^{E_{g-1}} dE \;
|
||||
\sigma_X (r, E) \psi (r, E, \Omega)}{\int_{r \in V} dr \int_{4\pi}
|
||||
d\Omega \int_{E_g}^{E_{g-1}} dE \; \psi (r, E, \Omega)}
|
||||
|
||||
where :math:`\sigma_X` is the requested reaction type of interest.
|
||||
|
||||
Parameters
|
||||
----------
|
||||
rxn_type : str
|
||||
Reaction type (e.g., '(n,2n)', '(n,Xt)', etc.)
|
||||
domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.RegularMesh
|
||||
The domain for spatial homogenization
|
||||
domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
|
||||
The domain type for spatial homogenization
|
||||
groups : openmc.mgxs.EnergyGroups
|
||||
The energy group structure for energy condensation
|
||||
by_nuclide : bool
|
||||
If true, computes cross sections for each nuclide in domain
|
||||
name : str, optional
|
||||
Name of the multi-group cross section. Used as a label to identify
|
||||
tallies in OpenMC 'tallies.xml' file.
|
||||
num_polar : Integral, optional
|
||||
Number of equi-width polar angle bins for angle discretization;
|
||||
defaults to one bin
|
||||
num_azimuthal : Integral, optional
|
||||
Number of equi-width azimuthal angle bins for angle discretization;
|
||||
defaults to one bin
|
||||
|
||||
Attributes
|
||||
----------
|
||||
name : str, optional
|
||||
Name of the multi-group cross section
|
||||
rxn_type : str
|
||||
Reaction type (e.g., '(n,2n)', '(n,Xt)', etc.)
|
||||
by_nuclide : bool
|
||||
If true, computes cross sections for each nuclide in domain
|
||||
domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.RegularMesh
|
||||
Domain for spatial homogenization
|
||||
domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
|
||||
Domain type for spatial homogenization
|
||||
energy_groups : openmc.mgxs.EnergyGroups
|
||||
Energy group structure for energy condensation
|
||||
num_polar : Integral
|
||||
Number of equi-width polar angle bins for angle discretization
|
||||
num_azimuthal : Integral
|
||||
Number of equi-width azimuthal angle bins for angle discretization
|
||||
tally_trigger : openmc.Trigger
|
||||
An (optional) tally precision trigger given to each tally used to
|
||||
compute the cross section
|
||||
scores : list of str
|
||||
The scores in each tally used to compute the multi-group cross section
|
||||
filters : list of openmc.Filter
|
||||
The filters in each tally used to compute the multi-group cross section
|
||||
tally_keys : list of str
|
||||
The keys into the tallies dictionary for each tally used to compute
|
||||
the multi-group cross section
|
||||
estimator : {'tracklength', 'collision', 'analog'}
|
||||
The tally estimator used to compute the multi-group cross section
|
||||
tallies : collections.OrderedDict
|
||||
OpenMC tallies needed to compute the multi-group cross section. The keys
|
||||
are strings listed in the :attr:`TotalXS.tally_keys` property and values
|
||||
are instances of :class:`openmc.Tally`.
|
||||
rxn_rate_tally : openmc.Tally
|
||||
Derived tally for the reaction rate tally used in the numerator to
|
||||
compute the multi-group cross section. This attribute is None
|
||||
unless the multi-group cross section has been computed.
|
||||
xs_tally : openmc.Tally
|
||||
Derived tally for the multi-group cross section. This attribute
|
||||
is None unless the multi-group cross section has been computed.
|
||||
num_subdomains : int
|
||||
The number of subdomains is unity for 'material', 'cell' and 'universe'
|
||||
domain types. This is equal to the number of cell instances
|
||||
for 'distribcell' domain types (it is equal to unity prior to loading
|
||||
tally data from a statepoint file).
|
||||
num_nuclides : int
|
||||
The number of nuclides for which the multi-group cross section is
|
||||
being tracked. This is unity if the by_nuclide attribute is False.
|
||||
nuclides : Iterable of str or 'sum'
|
||||
The optional user-specified nuclides for which to compute cross
|
||||
sections (e.g., 'U238', 'O16'). If by_nuclide is True but nuclides
|
||||
are not specified by the user, all nuclides in the spatial domain
|
||||
are included. This attribute is 'sum' if by_nuclide is false.
|
||||
sparse : bool
|
||||
Whether or not the MGXS' tallies use SciPy's LIL sparse matrix format
|
||||
for compressed data storage
|
||||
loaded_sp : bool
|
||||
Whether or not a statepoint file has been loaded with tally data
|
||||
derived : bool
|
||||
Whether or not the MGXS is merged from one or more other MGXS
|
||||
hdf5_key : str
|
||||
The key used to index multi-group cross sections in an HDF5 data store
|
||||
|
||||
"""
|
||||
|
||||
def __init__(self, rxn_type, domain=None, domain_type=None, groups=None,
|
||||
by_nuclide=False, name='', num_polar=1, num_azimuthal=1):
|
||||
cv.check_value("rxn_type", rxn_type, ARBITRARY_VECTOR_TYPES)
|
||||
super().__init__(domain, domain_type, groups, by_nuclide, name,
|
||||
num_polar, num_azimuthal)
|
||||
self._rxn_type = rxn_type
|
||||
|
||||
|
||||
class ArbitraryMatrixXS(MatrixMGXS):
|
||||
r"""A multi-group matrix cross section for an arbitrary reaction type.
|
||||
|
||||
This class can be used for both OpenMC input generation and tally data
|
||||
post-processing to compute spatially-homogenized and energy-integrated
|
||||
multi-group cross sections for multi-group neutronics calculations. At a
|
||||
minimum, one needs to set the :attr:`ArbitraryMatrixXS.energy_groups` and
|
||||
:attr:`ArbitraryMatrixXS.domain` properties. Tallies for the flux and
|
||||
appropriate reaction rates over the specified domain are generated
|
||||
automatically via the :attr:`ArbitraryMatrixXS.tallies` property, which can
|
||||
then be appended to a :class:`openmc.Tallies` instance.
|
||||
|
||||
For post-processing, the :meth:`MGXS.load_from_statepoint` will pull in the
|
||||
necessary data to compute multi-group cross sections from a
|
||||
:class:`openmc.StatePoint` instance. The derived multi-group cross section
|
||||
can then be obtained from the :attr:`ArbitraryMatrixXS.xs_tally` property.
|
||||
|
||||
For a spatial domain :math:`V`, incoming energy group
|
||||
:math:`[E_{g'},E_{g'-1}]`, and outgoing energy group :math:`[E_g,E_{g-1}]`,
|
||||
the fission production is calculated as:
|
||||
|
||||
.. math::
|
||||
|
||||
\begin{aligned}
|
||||
\langle \sigma_{X,g'\rightarrow g} \phi \rangle &= \int_{r \in V} dr
|
||||
\int_{4\pi} d\Omega' \int_{E_{g'}}^{E_{g'-1}} dE' \int_{E_g}^{E_{g-1}} dE
|
||||
\; \chi(E) \sigma_X (r, E') \psi(r, E', \Omega')\\
|
||||
\langle \phi \rangle &= \int_{r \in V} dr \int_{4\pi} d\Omega
|
||||
\int_{E_g}^{E_{g-1}} dE \; \psi (r, E, \Omega) \\
|
||||
\sigma_{X,g'\rightarrow g} &= \frac{\langle \sigma_{X,g'\rightarrow
|
||||
g} \phi \rangle}{\langle \phi \rangle}
|
||||
\end{aligned}
|
||||
|
||||
where :math:`\sigma_X` is the requested reaction type of interest.
|
||||
|
||||
Parameters
|
||||
----------
|
||||
rxn_type : str
|
||||
Reaction type (e.g., '(n,2n)', '(n,nta)', etc.). Valid names have
|
||||
neutrons as a product.
|
||||
domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.RegularMesh
|
||||
The domain for spatial homogenization
|
||||
domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
|
||||
The domain type for spatial homogenization
|
||||
groups : openmc.mgxs.EnergyGroups
|
||||
The energy group structure for energy condensation
|
||||
by_nuclide : bool
|
||||
If true, computes cross sections for each nuclide in domain
|
||||
name : str, optional
|
||||
Name of the multi-group cross section. Used as a label to identify
|
||||
tallies in OpenMC 'tallies.xml' file.
|
||||
num_polar : Integral, optional
|
||||
Number of equi-width polar angle bins for angle discretization;
|
||||
defaults to one bin
|
||||
num_azimuthal : Integral, optional
|
||||
Number of equi-width azimuthal angle bins for angle discretization;
|
||||
defaults to one bin
|
||||
|
||||
Attributes
|
||||
----------
|
||||
name : str, optional
|
||||
Name of the multi-group cross section
|
||||
rxn_type : str
|
||||
Reaction type (e.g., 'total', 'nu-fission', etc.)
|
||||
by_nuclide : bool
|
||||
If true, computes cross sections for each nuclide in domain
|
||||
domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.RegularMesh
|
||||
Domain for spatial homogenization
|
||||
domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
|
||||
Domain type for spatial homogenization
|
||||
energy_groups : openmc.mgxs.EnergyGroups
|
||||
Energy group structure for energy condensation
|
||||
num_polar : Integral
|
||||
Number of equi-width polar angle bins for angle discretization
|
||||
num_azimuthal : Integral
|
||||
Number of equi-width azimuthal angle bins for angle discretization
|
||||
tally_trigger : openmc.Trigger
|
||||
An (optional) tally precision trigger given to each tally used to
|
||||
compute the cross section
|
||||
scores : list of str
|
||||
The scores in each tally used to compute the multi-group cross section
|
||||
filters : list of openmc.Filter
|
||||
The filters in each tally used to compute the multi-group cross section
|
||||
tally_keys : list of str
|
||||
The keys into the tallies dictionary for each tally used to compute
|
||||
the multi-group cross section
|
||||
estimator : 'analog'
|
||||
The tally estimator used to compute the multi-group cross section
|
||||
tallies : collections.OrderedDict
|
||||
OpenMC tallies needed to compute the multi-group cross section. The keys
|
||||
are strings listed in the :attr:`NuFissionMatrixXS.tally_keys`
|
||||
property and values are instances of :class:`openmc.Tally`.
|
||||
rxn_rate_tally : openmc.Tally
|
||||
Derived tally for the reaction rate tally used in the numerator to
|
||||
compute the multi-group cross section. This attribute is None
|
||||
unless the multi-group cross section has been computed.
|
||||
xs_tally : openmc.Tally
|
||||
Derived tally for the multi-group cross section. This attribute
|
||||
is None unless the multi-group cross section has been computed.
|
||||
num_subdomains : int
|
||||
The number of subdomains is unity for 'material', 'cell' and 'universe'
|
||||
domain types. This is equal to the number of cell instances
|
||||
for 'distribcell' domain types (it is equal to unity prior to loading
|
||||
tally data from a statepoint file).
|
||||
num_nuclides : int
|
||||
The number of nuclides for which the multi-group cross section is
|
||||
being tracked. This is unity if the by_nuclide attribute is False.
|
||||
nuclides : Iterable of str or 'sum'
|
||||
The optional user-specified nuclides for which to compute cross
|
||||
sections (e.g., 'U238', 'O16'). If by_nuclide is True but nuclides
|
||||
are not specified by the user, all nuclides in the spatial domain
|
||||
are included. This attribute is 'sum' if by_nuclide is false.
|
||||
sparse : bool
|
||||
Whether or not the MGXS' tallies use SciPy's LIL sparse matrix format
|
||||
for compressed data storage
|
||||
loaded_sp : bool
|
||||
Whether or not a statepoint file has been loaded with tally data
|
||||
derived : bool
|
||||
Whether or not the MGXS is merged from one or more other MGXS
|
||||
hdf5_key : str
|
||||
The key used to index multi-group cross sections in an HDF5 data store
|
||||
|
||||
"""
|
||||
|
||||
def __init__(self, rxn_type, domain=None, domain_type=None, groups=None,
|
||||
by_nuclide=False, name='', num_polar=1,
|
||||
num_azimuthal=1):
|
||||
cv.check_value("rxn_type", rxn_type, ARBITRARY_MATRIX_TYPES)
|
||||
super().__init__(domain, domain_type, groups, by_nuclide, name,
|
||||
num_polar, num_azimuthal)
|
||||
self._rxn_type = rxn_type.split(" ")[0]
|
||||
self._estimator = 'analog'
|
||||
self._valid_estimators = ['analog']
|
||||
|
||||
|
||||
class ScatterMatrixXS(MatrixMGXS):
|
||||
r"""A scattering matrix multi-group cross section with the cosine of the
|
||||
change-in-angle represented as one or more Legendre moments or a histogram.
|
||||
|
|
|
|||
|
|
@ -243,7 +243,7 @@ std::unordered_map<int, std::string> REACTION_NAME_MAP {
|
|||
{N_3N3HE, "(n,3n3He)"},
|
||||
{N_4N3HE, "(n,4n3He)"},
|
||||
{N_3N2P, "(n,3n2p)"},
|
||||
{N_3N3A, "(n,3n3a)"},
|
||||
{N_3N2A, "(n,3n2a)"},
|
||||
{N_3NPA, "(n,3npa)"},
|
||||
{N_DT, "(n,dt)"},
|
||||
{N_NPD, "(n,npd)"},
|
||||
|
|
|
|||
File diff suppressed because it is too large
Load diff
|
|
@ -1,33 +1,44 @@
|
|||
total
|
||||
material group in nuclide mean std. dev.
|
||||
1 1 1 total 0.414825 0.022793
|
||||
0 1 2 total 0.660170 0.047519
|
||||
transport
|
||||
material group in nuclide mean std. dev.
|
||||
1 1 1 total 0.363092 0.023838
|
||||
0 1 2 total 0.644851 0.047675
|
||||
nu-transport
|
||||
material group in nuclide mean std. dev.
|
||||
1 1 1 total 0.363092 0.023838
|
||||
0 1 2 total 0.644851 0.047675
|
||||
absorption
|
||||
material group in nuclide mean std. dev.
|
||||
1 1 1 total 0.027408 0.002692
|
||||
0 1 2 total 0.264511 0.023367
|
||||
capture
|
||||
material group in nuclide mean std. dev.
|
||||
1 1 1 total 0.019845 0.002643
|
||||
0 1 2 total 0.071719 0.025208
|
||||
fission
|
||||
material group in nuclide mean std. dev.
|
||||
1 1 1 total 0.007563 0.000508
|
||||
0 1 2 total 0.192791 0.017106
|
||||
nu-fission
|
||||
material group in nuclide mean std. dev.
|
||||
1 1 1 total 0.019432 0.001323
|
||||
0 1 2 total 0.469775 0.041682
|
||||
kappa-fission
|
||||
material group in nuclide mean std. dev.
|
||||
1 1 1 total 1.474570e+06 9.923536e+04
|
||||
0 1 2 total 3.728689e+07 3.308375e+06
|
||||
scatter
|
||||
material group in nuclide mean std. dev.
|
||||
1 1 1 total 0.387418 0.020626
|
||||
0 1 2 total 0.395659 0.025125
|
||||
nu-scatter
|
||||
material group in nuclide mean std. dev.
|
||||
1 1 1 total 0.385188 0.026946
|
||||
0 1 2 total 0.412389 0.015425
|
||||
scatter matrix
|
||||
material group in group out legendre nuclide mean std. dev.
|
||||
12 1 1 1 P0 total 0.384199 0.027001
|
||||
13 1 1 1 P1 total 0.051870 0.006983
|
||||
|
|
@ -45,6 +56,7 @@
|
|||
1 1 2 2 P1 total 0.016482 0.004502
|
||||
2 1 2 2 P2 total 0.006371 0.010551
|
||||
3 1 2 2 P3 total -0.010499 0.010438
|
||||
nu-scatter matrix
|
||||
material group in group out legendre nuclide mean std. dev.
|
||||
12 1 1 1 P0 total 0.384199 0.027001
|
||||
13 1 1 1 P1 total 0.051870 0.006983
|
||||
|
|
@ -62,21 +74,25 @@
|
|||
1 1 2 2 P1 total 0.016482 0.004502
|
||||
2 1 2 2 P2 total 0.006371 0.010551
|
||||
3 1 2 2 P3 total -0.010499 0.010438
|
||||
multiplicity matrix
|
||||
material group in group out nuclide mean std. dev.
|
||||
3 1 1 1 total 1.0 0.078516
|
||||
2 1 1 2 total 1.0 0.687184
|
||||
1 1 2 1 total 1.0 1.414214
|
||||
0 1 2 2 total 1.0 0.041130
|
||||
nu-fission matrix
|
||||
material group in group out nuclide mean std. dev.
|
||||
3 1 1 1 total 0.020142 0.003149
|
||||
2 1 1 2 total 0.000000 0.000000
|
||||
1 1 2 1 total 0.454366 0.027426
|
||||
0 1 2 2 total 0.000000 0.000000
|
||||
scatter probability matrix
|
||||
material group in group out nuclide mean std. dev.
|
||||
3 1 1 1 total 0.997433 0.078224
|
||||
2 1 1 2 total 0.002567 0.001256
|
||||
1 1 2 1 total 0.002242 0.002243
|
||||
0 1 2 2 total 0.997758 0.041053
|
||||
consistent scatter matrix
|
||||
material group in group out legendre nuclide mean std. dev.
|
||||
12 1 1 1 P0 total 0.386423 0.036629
|
||||
13 1 1 1 P1 total 0.052170 0.007767
|
||||
|
|
@ -94,6 +110,7 @@
|
|||
1 1 2 2 P1 total 0.015813 0.004443
|
||||
2 1 2 2 P2 total 0.006113 0.010131
|
||||
3 1 2 2 P3 total -0.010073 0.010037
|
||||
consistent nu-scatter matrix
|
||||
material group in group out legendre nuclide mean std. dev.
|
||||
12 1 1 1 P0 total 0.386423 0.047563
|
||||
13 1 1 1 P1 total 0.052170 0.008781
|
||||
|
|
@ -111,33 +128,107 @@
|
|||
1 1 2 2 P1 total 0.015813 0.004491
|
||||
2 1 2 2 P2 total 0.006113 0.010134
|
||||
3 1 2 2 P3 total -0.010073 0.010045
|
||||
chi
|
||||
material group out nuclide mean std. dev.
|
||||
1 1 1 total 1.0 0.046071
|
||||
0 1 2 total 0.0 0.000000
|
||||
chi-prompt
|
||||
material group out nuclide mean std. dev.
|
||||
1 1 1 total 1.0 0.051471
|
||||
0 1 2 total 0.0 0.000000
|
||||
inverse-velocity
|
||||
material group in nuclide mean std. dev.
|
||||
1 1 1 total 5.709324e-08 4.687938e-09
|
||||
0 1 2 total 2.855739e-06 2.442164e-07
|
||||
prompt-nu-fission
|
||||
material group in nuclide mean std. dev.
|
||||
1 1 1 total 0.019239 0.001310
|
||||
0 1 2 total 0.466719 0.041411
|
||||
prompt-nu-fission matrix
|
||||
material group in group out nuclide mean std. dev.
|
||||
3 1 1 1 total 0.020142 0.003149
|
||||
2 1 1 2 total 0.000000 0.000000
|
||||
1 1 2 1 total 0.445819 0.028675
|
||||
0 1 2 2 total 0.000000 0.000000
|
||||
diffusion-coefficient
|
||||
material group in legendre nuclide mean std. dev.
|
||||
2 1 1 P0 total 10.942878 12.704137
|
||||
3 1 1 P1 total 0.918041 0.075837
|
||||
0 1 2 P0 total 1.370855 0.292448
|
||||
1 1 2 P1 total 0.516916 0.050251
|
||||
nu-diffusion-coefficient
|
||||
material group in legendre nuclide mean std. dev.
|
||||
2 1 1 P0 total 10.942878 12.704137
|
||||
3 1 1 P1 total 0.918041 0.075837
|
||||
0 1 2 P0 total 1.370855 0.292448
|
||||
1 1 2 P1 total 0.516916 0.050251
|
||||
(n,elastic)
|
||||
material group in nuclide mean std. dev.
|
||||
1 1 1 total 0.358207 0.019662
|
||||
0 1 2 total 0.395659 0.025125
|
||||
(n,level)
|
||||
material group in nuclide mean std. dev.
|
||||
1 1 1 total 0.000619 0.000049
|
||||
0 1 2 total 0.000000 0.000000
|
||||
(n,2n)
|
||||
material group in nuclide mean std. dev.
|
||||
1 1 1 total 0.000121 0.000054
|
||||
0 1 2 total 0.000000 0.000000
|
||||
(n,na)
|
||||
material group in nuclide mean std. dev.
|
||||
1 1 1 total 9.209364e-11 9.154237e-11
|
||||
0 1 2 total 0.000000e+00 0.000000e+00
|
||||
(n,nc)
|
||||
material group in nuclide mean std. dev.
|
||||
1 1 1 total 0.009458 0.000768
|
||||
0 1 2 total 0.000000 0.000000
|
||||
(n,gamma)
|
||||
material group in nuclide mean std. dev.
|
||||
1 1 1 total 0.019727 0.002262
|
||||
0 1 2 total 0.071719 0.006262
|
||||
(n,a)
|
||||
material group in nuclide mean std. dev.
|
||||
1 1 1 total 0.000124 0.000023
|
||||
0 1 2 total 0.000000 0.000000
|
||||
(n,Xa)
|
||||
material group in nuclide mean std. dev.
|
||||
1 1 1 total 0.000124 0.000023
|
||||
0 1 2 total 0.000000 0.000000
|
||||
heating
|
||||
material group in nuclide mean std. dev.
|
||||
1 1 1 total 1.287977e+06 8.806399e+04
|
||||
0 1 2 total 3.222710e+07 2.903396e+06
|
||||
damage-energy
|
||||
material group in nuclide mean std. dev.
|
||||
1 1 1 total 2471.829371 114.012620
|
||||
0 1 2 total 1357.269536 120.427363
|
||||
(n,n1)
|
||||
material group in nuclide mean std. dev.
|
||||
1 1 1 total 0.011877 0.000507
|
||||
0 1 2 total 0.000000 0.000000
|
||||
(n,a0)
|
||||
material group in nuclide mean std. dev.
|
||||
1 1 1 total 0.000115 0.000024
|
||||
0 1 2 total 0.000000 0.000000
|
||||
(n,nc) matrix
|
||||
material group in group out nuclide mean std. dev.
|
||||
3 1 1 1 total 0.010878 0.001184
|
||||
2 1 1 2 total 0.000000 0.000000
|
||||
1 1 2 1 total 0.000000 0.000000
|
||||
0 1 2 2 total 0.000000 0.000000
|
||||
(n,n1) matrix
|
||||
material group in group out nuclide mean std. dev.
|
||||
3 1 1 1 total 0.011043 0.000937
|
||||
2 1 1 2 total 0.000000 0.000000
|
||||
1 1 2 1 total 0.000000 0.000000
|
||||
0 1 2 2 total 0.000000 0.000000
|
||||
(n,2n) matrix
|
||||
material group in group out nuclide mean std. dev.
|
||||
3 1 1 1 total 0.0 0.0
|
||||
2 1 1 2 total 0.0 0.0
|
||||
1 1 2 1 total 0.0 0.0
|
||||
0 1 2 2 total 0.0 0.0
|
||||
delayed-nu-fission
|
||||
material delayedgroup group in nuclide mean std. dev.
|
||||
1 1 1 1 total 0.000004 2.897486e-07
|
||||
3 1 2 1 total 0.000027 1.850037e-06
|
||||
|
|
@ -151,6 +242,7 @@
|
|||
6 1 4 2 total 0.001182 1.048679e-04
|
||||
8 1 5 2 total 0.000485 4.299445e-05
|
||||
10 1 6 2 total 0.000203 1.801022e-05
|
||||
chi-delayed
|
||||
material delayedgroup group out nuclide mean std. dev.
|
||||
1 1 1 1 total 0.0 0.000000
|
||||
3 1 2 1 total 1.0 0.869128
|
||||
|
|
@ -164,6 +256,7 @@
|
|||
6 1 4 2 total 0.0 0.000000
|
||||
8 1 5 2 total 0.0 0.000000
|
||||
10 1 6 2 total 0.0 0.000000
|
||||
beta
|
||||
material delayedgroup group in nuclide mean std. dev.
|
||||
1 1 1 1 total 0.000222 0.000018
|
||||
3 1 2 1 total 0.001388 0.000115
|
||||
|
|
@ -177,6 +270,7 @@
|
|||
6 1 4 2 total 0.002516 0.000273
|
||||
8 1 5 2 total 0.001031 0.000112
|
||||
10 1 6 2 total 0.000432 0.000047
|
||||
decay-rate
|
||||
material delayedgroup nuclide mean std. dev.
|
||||
0 1 1 total 0.013355 0.001272
|
||||
1 1 2 total 0.032600 0.003048
|
||||
|
|
@ -184,6 +278,7 @@
|
|||
3 1 4 total 0.305910 0.027728
|
||||
4 1 5 total 0.861934 0.075425
|
||||
5 1 6 total 2.895065 0.253942
|
||||
delayed-nu-fission matrix
|
||||
material delayedgroup group in group out nuclide mean std. dev.
|
||||
3 1 1 1 1 total 0.000000 0.000000
|
||||
7 1 2 1 1 total 0.000000 0.000000
|
||||
|
|
@ -209,36 +304,47 @@
|
|||
12 1 4 2 2 total 0.000000 0.000000
|
||||
16 1 5 2 2 total 0.000000 0.000000
|
||||
20 1 6 2 2 total 0.000000 0.000000
|
||||
total
|
||||
material group in nuclide mean std. dev.
|
||||
1 2 1 total 0.313738 0.015582
|
||||
0 2 2 total 0.300821 0.028052
|
||||
transport
|
||||
material group in nuclide mean std. dev.
|
||||
1 2 1 total 0.275508 0.017742
|
||||
0 2 2 total 0.312035 0.032384
|
||||
nu-transport
|
||||
material group in nuclide mean std. dev.
|
||||
1 2 1 total 0.275508 0.017742
|
||||
0 2 2 total 0.312035 0.032384
|
||||
absorption
|
||||
material group in nuclide mean std. dev.
|
||||
1 2 1 total 0.001575 0.000323
|
||||
0 2 2 total 0.005400 0.000618
|
||||
capture
|
||||
material group in nuclide mean std. dev.
|
||||
1 2 1 total 0.001575 0.000323
|
||||
0 2 2 total 0.005400 0.000618
|
||||
fission
|
||||
material group in nuclide mean std. dev.
|
||||
1 2 1 total 0.0 0.0
|
||||
0 2 2 total 0.0 0.0
|
||||
nu-fission
|
||||
material group in nuclide mean std. dev.
|
||||
1 2 1 total 0.0 0.0
|
||||
0 2 2 total 0.0 0.0
|
||||
kappa-fission
|
||||
material group in nuclide mean std. dev.
|
||||
1 2 1 total 0.0 0.0
|
||||
0 2 2 total 0.0 0.0
|
||||
scatter
|
||||
material group in nuclide mean std. dev.
|
||||
1 2 1 total 0.312163 0.015322
|
||||
0 2 2 total 0.295421 0.027446
|
||||
nu-scatter
|
||||
material group in nuclide mean std. dev.
|
||||
1 2 1 total 0.310121 0.033788
|
||||
0 2 2 total 0.296264 0.043792
|
||||
scatter matrix
|
||||
material group in group out legendre nuclide mean std. dev.
|
||||
12 2 1 1 P0 total 0.310121 0.033788
|
||||
13 2 1 1 P1 total 0.038230 0.008484
|
||||
|
|
@ -256,6 +362,7 @@
|
|||
1 2 2 2 P1 total -0.011214 0.016180
|
||||
2 2 2 2 P2 total 0.008837 0.011504
|
||||
3 2 2 2 P3 total -0.003270 0.007329
|
||||
nu-scatter matrix
|
||||
material group in group out legendre nuclide mean std. dev.
|
||||
12 2 1 1 P0 total 0.310121 0.033788
|
||||
13 2 1 1 P1 total 0.038230 0.008484
|
||||
|
|
@ -273,21 +380,25 @@
|
|||
1 2 2 2 P1 total -0.011214 0.016180
|
||||
2 2 2 2 P2 total 0.008837 0.011504
|
||||
3 2 2 2 P3 total -0.003270 0.007329
|
||||
multiplicity matrix
|
||||
material group in group out nuclide mean std. dev.
|
||||
3 2 1 1 total 1.0 0.108779
|
||||
2 2 1 2 total 0.0 0.000000
|
||||
1 2 2 1 total 0.0 0.000000
|
||||
0 2 2 2 total 1.0 0.142427
|
||||
nu-fission matrix
|
||||
material group in group out nuclide mean std. dev.
|
||||
3 2 1 1 total 0.0 0.0
|
||||
2 2 1 2 total 0.0 0.0
|
||||
1 2 2 1 total 0.0 0.0
|
||||
0 2 2 2 total 0.0 0.0
|
||||
scatter probability matrix
|
||||
material group in group out nuclide mean std. dev.
|
||||
3 2 1 1 total 1.0 0.108779
|
||||
2 2 1 2 total 0.0 0.000000
|
||||
1 2 2 1 total 0.0 0.000000
|
||||
0 2 2 2 total 1.0 0.142427
|
||||
consistent scatter matrix
|
||||
material group in group out legendre nuclide mean std. dev.
|
||||
12 2 1 1 P0 total 0.312163 0.037253
|
||||
13 2 1 1 P1 total 0.038481 0.008743
|
||||
|
|
@ -305,6 +416,7 @@
|
|||
1 2 2 2 P1 total -0.011182 0.016162
|
||||
2 2 2 2 P2 total 0.008811 0.011495
|
||||
3 2 2 2 P3 total -0.003261 0.007313
|
||||
consistent nu-scatter matrix
|
||||
material group in group out legendre nuclide mean std. dev.
|
||||
12 2 1 1 P0 total 0.312163 0.050407
|
||||
13 2 1 1 P1 total 0.038481 0.009693
|
||||
|
|
@ -322,33 +434,107 @@
|
|||
1 2 2 2 P1 total -0.011182 0.016240
|
||||
2 2 2 2 P2 total 0.008811 0.011563
|
||||
3 2 2 2 P3 total -0.003261 0.007328
|
||||
chi
|
||||
material group out nuclide mean std. dev.
|
||||
1 2 1 total 0.0 0.0
|
||||
0 2 2 total 0.0 0.0
|
||||
chi-prompt
|
||||
material group out nuclide mean std. dev.
|
||||
1 2 1 total 0.0 0.0
|
||||
0 2 2 total 0.0 0.0
|
||||
inverse-velocity
|
||||
material group in nuclide mean std. dev.
|
||||
1 2 1 total 5.995979e-08 4.553085e-09
|
||||
0 2 2 total 2.985490e-06 3.417020e-07
|
||||
prompt-nu-fission
|
||||
material group in nuclide mean std. dev.
|
||||
1 2 1 total 0.0 0.0
|
||||
0 2 2 total 0.0 0.0
|
||||
prompt-nu-fission matrix
|
||||
material group in group out nuclide mean std. dev.
|
||||
3 2 1 1 total 0.0 0.0
|
||||
2 2 1 2 total 0.0 0.0
|
||||
1 2 2 1 total 0.0 0.0
|
||||
0 2 2 2 total 0.0 0.0
|
||||
diffusion-coefficient
|
||||
material group in legendre nuclide mean std. dev.
|
||||
2 2 1 P0 total 92.158584 948.055013
|
||||
3 2 1 P1 total 1.209886 0.097583
|
||||
0 2 2 P0 total 73.145434 834.775399
|
||||
1 2 2 P1 total 1.068256 0.148711
|
||||
nu-diffusion-coefficient
|
||||
material group in legendre nuclide mean std. dev.
|
||||
2 2 1 P0 total 92.158584 948.055013
|
||||
3 2 1 P1 total 1.209886 0.097583
|
||||
0 2 2 P0 total 73.145434 834.775399
|
||||
1 2 2 P1 total 1.068256 0.148711
|
||||
(n,elastic)
|
||||
material group in nuclide mean std. dev.
|
||||
1 2 1 total 0.301031 0.014977
|
||||
0 2 2 total 0.295421 0.027446
|
||||
(n,level)
|
||||
material group in nuclide mean std. dev.
|
||||
1 2 1 total 0.0 0.0
|
||||
0 2 2 total 0.0 0.0
|
||||
(n,2n)
|
||||
material group in nuclide mean std. dev.
|
||||
1 2 1 total 0.000005 0.000005
|
||||
0 2 2 total 0.000000 0.000000
|
||||
(n,na)
|
||||
material group in nuclide mean std. dev.
|
||||
1 2 1 total 1.259778e-09 1.152715e-09
|
||||
0 2 2 total 0.000000e+00 0.000000e+00
|
||||
(n,nc)
|
||||
material group in nuclide mean std. dev.
|
||||
1 2 1 total 0.002607 0.000356
|
||||
0 2 2 total 0.000000 0.000000
|
||||
(n,gamma)
|
||||
material group in nuclide mean std. dev.
|
||||
1 2 1 total 0.001569 0.000322
|
||||
0 2 2 total 0.005400 0.000618
|
||||
(n,a)
|
||||
material group in nuclide mean std. dev.
|
||||
1 2 1 total 1.032353e-06 1.308817e-07
|
||||
0 2 2 total 2.735898e-07 2.541726e-08
|
||||
(n,Xa)
|
||||
material group in nuclide mean std. dev.
|
||||
1 2 1 total 1.033613e-06 1.318182e-07
|
||||
0 2 2 total 2.735898e-07 2.541726e-08
|
||||
heating
|
||||
material group in nuclide mean std. dev.
|
||||
1 2 1 total 2819.023866 149.123603
|
||||
0 2 2 total 2.414215 0.196172
|
||||
damage-energy
|
||||
material group in nuclide mean std. dev.
|
||||
1 2 1 total 1712.186983 87.224250
|
||||
0 2 2 total 0.294202 0.032716
|
||||
(n,n1)
|
||||
material group in nuclide mean std. dev.
|
||||
1 2 1 total 0.003108 0.000252
|
||||
0 2 2 total 0.000000 0.000000
|
||||
(n,a0)
|
||||
material group in nuclide mean std. dev.
|
||||
1 2 1 total 8.426133e-07 7.088213e-08
|
||||
0 2 2 total 2.733143e-07 2.539166e-08
|
||||
(n,nc) matrix
|
||||
material group in group out nuclide mean std. dev.
|
||||
3 2 1 1 total 0.001782 0.000845
|
||||
2 2 1 2 total 0.000000 0.000000
|
||||
1 2 2 1 total 0.000000 0.000000
|
||||
0 2 2 2 total 0.000000 0.000000
|
||||
(n,n1) matrix
|
||||
material group in group out nuclide mean std. dev.
|
||||
3 2 1 1 total 0.002228 0.000725
|
||||
2 2 1 2 total 0.000000 0.000000
|
||||
1 2 2 1 total 0.000000 0.000000
|
||||
0 2 2 2 total 0.000000 0.000000
|
||||
(n,2n) matrix
|
||||
material group in group out nuclide mean std. dev.
|
||||
3 2 1 1 total 0.0 0.0
|
||||
2 2 1 2 total 0.0 0.0
|
||||
1 2 2 1 total 0.0 0.0
|
||||
0 2 2 2 total 0.0 0.0
|
||||
delayed-nu-fission
|
||||
material delayedgroup group in nuclide mean std. dev.
|
||||
1 2 1 1 total 0.0 0.0
|
||||
3 2 2 1 total 0.0 0.0
|
||||
|
|
@ -362,6 +548,7 @@
|
|||
6 2 4 2 total 0.0 0.0
|
||||
8 2 5 2 total 0.0 0.0
|
||||
10 2 6 2 total 0.0 0.0
|
||||
chi-delayed
|
||||
material delayedgroup group out nuclide mean std. dev.
|
||||
1 2 1 1 total 0.0 0.0
|
||||
3 2 2 1 total 0.0 0.0
|
||||
|
|
@ -375,6 +562,7 @@
|
|||
6 2 4 2 total 0.0 0.0
|
||||
8 2 5 2 total 0.0 0.0
|
||||
10 2 6 2 total 0.0 0.0
|
||||
beta
|
||||
material delayedgroup group in nuclide mean std. dev.
|
||||
1 2 1 1 total 0.0 0.0
|
||||
3 2 2 1 total 0.0 0.0
|
||||
|
|
@ -388,6 +576,7 @@
|
|||
6 2 4 2 total 0.0 0.0
|
||||
8 2 5 2 total 0.0 0.0
|
||||
10 2 6 2 total 0.0 0.0
|
||||
decay-rate
|
||||
material delayedgroup nuclide mean std. dev.
|
||||
0 2 1 total 0.0 0.0
|
||||
1 2 2 total 0.0 0.0
|
||||
|
|
@ -395,6 +584,7 @@
|
|||
3 2 4 total 0.0 0.0
|
||||
4 2 5 total 0.0 0.0
|
||||
5 2 6 total 0.0 0.0
|
||||
delayed-nu-fission matrix
|
||||
material delayedgroup group in group out nuclide mean std. dev.
|
||||
3 2 1 1 1 total 0.0 0.0
|
||||
7 2 2 1 1 total 0.0 0.0
|
||||
|
|
@ -420,36 +610,47 @@
|
|||
12 2 4 2 2 total 0.0 0.0
|
||||
16 2 5 2 2 total 0.0 0.0
|
||||
20 2 6 2 2 total 0.0 0.0
|
||||
total
|
||||
material group in nuclide mean std. dev.
|
||||
1 3 1 total 0.664572 0.031215
|
||||
0 3 2 total 2.052384 0.224343
|
||||
transport
|
||||
material group in nuclide mean std. dev.
|
||||
1 3 1 total 0.283323 0.035206
|
||||
0 3 2 total 1.499740 0.230902
|
||||
nu-transport
|
||||
material group in nuclide mean std. dev.
|
||||
1 3 1 total 0.283323 0.035206
|
||||
0 3 2 total 1.499740 0.230902
|
||||
absorption
|
||||
material group in nuclide mean std. dev.
|
||||
1 3 1 total 0.000690 0.000044
|
||||
0 3 2 total 0.031687 0.003747
|
||||
capture
|
||||
material group in nuclide mean std. dev.
|
||||
1 3 1 total 0.000690 0.000044
|
||||
0 3 2 total 0.031687 0.003747
|
||||
fission
|
||||
material group in nuclide mean std. dev.
|
||||
1 3 1 total 0.0 0.0
|
||||
0 3 2 total 0.0 0.0
|
||||
nu-fission
|
||||
material group in nuclide mean std. dev.
|
||||
1 3 1 total 0.0 0.0
|
||||
0 3 2 total 0.0 0.0
|
||||
kappa-fission
|
||||
material group in nuclide mean std. dev.
|
||||
1 3 1 total 0.0 0.0
|
||||
0 3 2 total 0.0 0.0
|
||||
scatter
|
||||
material group in nuclide mean std. dev.
|
||||
1 3 1 total 0.663882 0.031173
|
||||
0 3 2 total 2.020697 0.220604
|
||||
nu-scatter
|
||||
material group in nuclide mean std. dev.
|
||||
1 3 1 total 0.671269 0.026186
|
||||
0 3 2 total 2.035388 0.258060
|
||||
scatter matrix
|
||||
material group in group out legendre nuclide mean std. dev.
|
||||
12 3 1 1 P0 total 0.639901 0.024709
|
||||
13 3 1 1 P1 total 0.381167 0.016243
|
||||
|
|
@ -467,6 +668,7 @@
|
|||
1 3 2 2 P1 total 0.509940 0.051236
|
||||
2 3 2 2 P2 total 0.111175 0.013020
|
||||
3 3 2 2 P3 total 0.024988 0.008312
|
||||
nu-scatter matrix
|
||||
material group in group out legendre nuclide mean std. dev.
|
||||
12 3 1 1 P0 total 0.639901 0.024709
|
||||
13 3 1 1 P1 total 0.381167 0.016243
|
||||
|
|
@ -484,21 +686,25 @@
|
|||
1 3 2 2 P1 total 0.509940 0.051236
|
||||
2 3 2 2 P2 total 0.111175 0.013020
|
||||
3 3 2 2 P3 total 0.024988 0.008312
|
||||
multiplicity matrix
|
||||
material group in group out nuclide mean std. dev.
|
||||
3 3 1 1 total 1.0 0.038609
|
||||
2 3 1 2 total 1.0 0.067667
|
||||
1 3 2 1 total 1.0 1.414214
|
||||
0 3 2 2 total 1.0 0.135929
|
||||
nu-fission matrix
|
||||
material group in group out nuclide mean std. dev.
|
||||
3 3 1 1 total 0.0 0.0
|
||||
2 3 1 2 total 0.0 0.0
|
||||
1 3 2 1 total 0.0 0.0
|
||||
0 3 2 2 total 0.0 0.0
|
||||
scatter probability matrix
|
||||
material group in group out nuclide mean std. dev.
|
||||
3 3 1 1 total 0.953271 0.036018
|
||||
2 3 1 2 total 0.046729 0.002547
|
||||
1 3 2 1 total 0.000218 0.000219
|
||||
0 3 2 2 total 0.999782 0.135885
|
||||
consistent scatter matrix
|
||||
material group in group out legendre nuclide mean std. dev.
|
||||
12 3 1 1 P0 total 0.632859 0.038142
|
||||
13 3 1 1 P1 total 0.376973 0.023715
|
||||
|
|
@ -516,6 +722,7 @@
|
|||
1 3 2 2 P1 total 0.506260 0.079140
|
||||
2 3 2 2 P2 total 0.110372 0.018488
|
||||
3 3 2 2 P3 total 0.024808 0.008771
|
||||
consistent nu-scatter matrix
|
||||
material group in group out legendre nuclide mean std. dev.
|
||||
12 3 1 1 P0 total 0.632859 0.045297
|
||||
13 3 1 1 P1 total 0.376973 0.027825
|
||||
|
|
@ -533,33 +740,107 @@
|
|||
1 3 2 2 P1 total 0.506260 0.104875
|
||||
2 3 2 2 P2 total 0.110372 0.023809
|
||||
3 3 2 2 P3 total 0.024808 0.009397
|
||||
chi
|
||||
material group out nuclide mean std. dev.
|
||||
1 3 1 total 0.0 0.0
|
||||
0 3 2 total 0.0 0.0
|
||||
chi-prompt
|
||||
material group out nuclide mean std. dev.
|
||||
1 3 1 total 0.0 0.0
|
||||
0 3 2 total 0.0 0.0
|
||||
inverse-velocity
|
||||
material group in nuclide mean std. dev.
|
||||
1 3 1 total 6.022078e-08 3.780437e-09
|
||||
0 3 2 total 3.044955e-06 3.600077e-07
|
||||
prompt-nu-fission
|
||||
material group in nuclide mean std. dev.
|
||||
1 3 1 total 0.0 0.0
|
||||
0 3 2 total 0.0 0.0
|
||||
prompt-nu-fission matrix
|
||||
material group in group out nuclide mean std. dev.
|
||||
3 3 1 1 total 0.0 0.0
|
||||
2 3 1 2 total 0.0 0.0
|
||||
1 3 2 1 total 0.0 0.0
|
||||
0 3 2 2 total 0.0 0.0
|
||||
diffusion-coefficient
|
||||
material group in legendre nuclide mean std. dev.
|
||||
2 3 1 P0 total 13.561252 21.988396
|
||||
3 3 1 P1 total 1.176515 0.154807
|
||||
0 3 2 P0 total -2.459765 6.393743
|
||||
1 3 2 P1 total 0.222261 0.040518
|
||||
nu-diffusion-coefficient
|
||||
material group in legendre nuclide mean std. dev.
|
||||
2 3 1 P0 total 13.561252 21.988396
|
||||
3 3 1 P1 total 1.176515 0.154807
|
||||
0 3 2 P0 total -2.459765 6.393743
|
||||
1 3 2 P1 total 0.222261 0.040518
|
||||
(n,elastic)
|
||||
material group in nuclide mean std. dev.
|
||||
1 3 1 total 0.663837 0.031175
|
||||
0 3 2 total 2.020697 0.220604
|
||||
(n,level)
|
||||
material group in nuclide mean std. dev.
|
||||
1 3 1 total 0.000045 0.000027
|
||||
0 3 2 total 0.000000 0.000000
|
||||
(n,2n)
|
||||
material group in nuclide mean std. dev.
|
||||
1 3 1 total 0.0 0.0
|
||||
0 3 2 total 0.0 0.0
|
||||
(n,na)
|
||||
material group in nuclide mean std. dev.
|
||||
1 3 1 total 6.743945e-11 6.604830e-11
|
||||
0 3 2 total 0.000000e+00 0.000000e+00
|
||||
(n,nc)
|
||||
material group in nuclide mean std. dev.
|
||||
1 3 1 total 0.0 0.0
|
||||
0 3 2 total 0.0 0.0
|
||||
(n,gamma)
|
||||
material group in nuclide mean std. dev.
|
||||
1 3 1 total 0.000219 0.000014
|
||||
0 3 2 total 0.011034 0.001305
|
||||
(n,a)
|
||||
material group in nuclide mean std. dev.
|
||||
1 3 1 total 0.000471 0.000031
|
||||
0 3 2 total 0.020653 0.002442
|
||||
(n,Xa)
|
||||
material group in nuclide mean std. dev.
|
||||
1 3 1 total 0.000472 0.000031
|
||||
0 3 2 total 0.020653 0.002442
|
||||
heating
|
||||
material group in nuclide mean std. dev.
|
||||
1 3 1 total 86854.045343 3785.266424
|
||||
0 3 2 total 61907.411870 6000.898789
|
||||
damage-energy
|
||||
material group in nuclide mean std. dev.
|
||||
1 3 1 total 1186.090822 44.862980
|
||||
0 3 2 total 337.187426 39.868517
|
||||
(n,n1)
|
||||
material group in nuclide mean std. dev.
|
||||
1 3 1 total 0.000001 6.118682e-07
|
||||
0 3 2 total 0.000000 0.000000e+00
|
||||
(n,a0)
|
||||
material group in nuclide mean std. dev.
|
||||
1 3 1 total 0.000084 0.000012
|
||||
0 3 2 total 0.001299 0.000154
|
||||
(n,nc) matrix
|
||||
material group in group out nuclide mean std. dev.
|
||||
3 3 1 1 total 0.0 0.0
|
||||
2 3 1 2 total 0.0 0.0
|
||||
1 3 2 1 total 0.0 0.0
|
||||
0 3 2 2 total 0.0 0.0
|
||||
(n,n1) matrix
|
||||
material group in group out nuclide mean std. dev.
|
||||
3 3 1 1 total 0.0 0.0
|
||||
2 3 1 2 total 0.0 0.0
|
||||
1 3 2 1 total 0.0 0.0
|
||||
0 3 2 2 total 0.0 0.0
|
||||
(n,2n) matrix
|
||||
material group in group out nuclide mean std. dev.
|
||||
3 3 1 1 total 0.0 0.0
|
||||
2 3 1 2 total 0.0 0.0
|
||||
1 3 2 1 total 0.0 0.0
|
||||
0 3 2 2 total 0.0 0.0
|
||||
delayed-nu-fission
|
||||
material delayedgroup group in nuclide mean std. dev.
|
||||
1 3 1 1 total 0.0 0.0
|
||||
3 3 2 1 total 0.0 0.0
|
||||
|
|
@ -573,6 +854,7 @@
|
|||
6 3 4 2 total 0.0 0.0
|
||||
8 3 5 2 total 0.0 0.0
|
||||
10 3 6 2 total 0.0 0.0
|
||||
chi-delayed
|
||||
material delayedgroup group out nuclide mean std. dev.
|
||||
1 3 1 1 total 0.0 0.0
|
||||
3 3 2 1 total 0.0 0.0
|
||||
|
|
@ -586,6 +868,7 @@
|
|||
6 3 4 2 total 0.0 0.0
|
||||
8 3 5 2 total 0.0 0.0
|
||||
10 3 6 2 total 0.0 0.0
|
||||
beta
|
||||
material delayedgroup group in nuclide mean std. dev.
|
||||
1 3 1 1 total 0.0 0.0
|
||||
3 3 2 1 total 0.0 0.0
|
||||
|
|
@ -599,6 +882,7 @@
|
|||
6 3 4 2 total 0.0 0.0
|
||||
8 3 5 2 total 0.0 0.0
|
||||
10 3 6 2 total 0.0 0.0
|
||||
decay-rate
|
||||
material delayedgroup nuclide mean std. dev.
|
||||
0 3 1 total 0.0 0.0
|
||||
1 3 2 total 0.0 0.0
|
||||
|
|
@ -606,6 +890,7 @@
|
|||
3 3 4 total 0.0 0.0
|
||||
4 3 5 total 0.0 0.0
|
||||
5 3 6 total 0.0 0.0
|
||||
delayed-nu-fission matrix
|
||||
material delayedgroup group in group out nuclide mean std. dev.
|
||||
3 3 1 1 1 total 0.0 0.0
|
||||
7 3 2 1 1 total 0.0 0.0
|
||||
|
|
|
|||
|
|
@ -19,9 +19,17 @@ class MGXSTestHarness(PyAPITestHarness):
|
|||
self.mgxs_lib = openmc.mgxs.Library(self._model.geometry)
|
||||
self.mgxs_lib.by_nuclide = False
|
||||
|
||||
# Test all relevant MGXS types
|
||||
# Test relevant MGXS types
|
||||
relevant_MGXS_TYPES = [item for item in openmc.mgxs.MGXS_TYPES
|
||||
if item != 'current']
|
||||
# Add in a subset of openmc.mgxs.ARBITRARY_VECTOR_TYPES and
|
||||
# openmc.mgxs.ARBITRARY_MATRIX_TYPES so we can see the code works,
|
||||
# but not use too much resources
|
||||
relevant_MGXS_TYPES += [
|
||||
"(n,elastic)", "(n,level)", "(n,2n)", "(n,na)", "(n,nc)",
|
||||
"(n,gamma)", "(n,a)", "(n,Xa)", "heating", "damage-energy",
|
||||
"(n,n1)", "(n,a0)", "(n,nc) matrix", "(n,n1) matrix",
|
||||
"(n,2n) matrix"]
|
||||
self.mgxs_lib.mgxs_types = tuple(relevant_MGXS_TYPES) + \
|
||||
openmc.mgxs.MDGXS_TYPES
|
||||
self.mgxs_lib.energy_groups = energy_groups
|
||||
|
|
@ -48,7 +56,7 @@ class MGXSTestHarness(PyAPITestHarness):
|
|||
for mgxs_type in self.mgxs_lib.mgxs_types:
|
||||
mgxs = self.mgxs_lib.get_mgxs(domain, mgxs_type)
|
||||
df = mgxs.get_pandas_dataframe()
|
||||
outstr += df.to_string() + '\n'
|
||||
outstr += mgxs_type + '\n' + df.to_string() + '\n'
|
||||
|
||||
# Hash the results if necessary
|
||||
if hash_output:
|
||||
|
|
|
|||
File diff suppressed because it is too large
Load diff
|
|
@ -1 +1 @@
|
|||
14f7fb2e399a4ad124b25b88dda5104d43cca9192aa1c17261da40bcc0e4394e56781f42f79c90107d0ff6b3e901f034bed09b037670e60b9ac4009b3805e17d
|
||||
a9999488e2aa2ad0f1d694afedb71fbb56f1d0c8e8abdb6616698505a9d85302bf90586866fdb58eac96f76712a2dfbc884cc03087df7b5d148fb29d14dd2bbb
|
||||
|
|
@ -18,9 +18,17 @@ class MGXSTestHarness(PyAPITestHarness):
|
|||
self.mgxs_lib = openmc.mgxs.Library(self._model.geometry)
|
||||
self.mgxs_lib.by_nuclide = True
|
||||
|
||||
# Test relevant all MGXS types
|
||||
# Test relevant MGXS types
|
||||
relevant_MGXS_TYPES = [item for item in openmc.mgxs.MGXS_TYPES
|
||||
if item != 'current']
|
||||
# Add in a subset of openmc.mgxs.ARBITRARY_VECTOR_TYPES and
|
||||
# openmc.mgxs.ARBITRARY_MATRIX_TYPES so we can see the code works,
|
||||
# but not use too much resources
|
||||
relevant_MGXS_TYPES += [
|
||||
"(n,elastic)", "(n,level)", "(n,2n)", "(n,na)", "(n,nc)",
|
||||
"(n,gamma)", "(n,a)", "(n,Xa)", "heating", "damage-energy",
|
||||
"(n,n1)", "(n,a0)", "(n,nc) matrix", "(n,n1) matrix",
|
||||
"(n,2n) matrix"]
|
||||
self.mgxs_lib.mgxs_types = tuple(relevant_MGXS_TYPES)
|
||||
self.mgxs_lib.energy_groups = energy_groups
|
||||
self.mgxs_lib.legendre_order = 3
|
||||
|
|
|
|||
Loading…
Add table
Add a link
Reference in a new issue