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Added nuclide and number density getter routines to Python API MultiGroupXS class

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This commit is contained in:
Will Boyd 2015-09-27 16:05:07 -04:00
parent fe16bb4b10
commit 0374a07c6f
4 changed files with 158 additions and 50 deletions
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4
openmc/cross.py
View file

@ -1,7 +1,7 @@
import sys
from openmc import Filter, Nuclide
from openmc.constants import FILTER_TYPES
from openmc.filter import _FILTER_TYPES
import openmc.checkvalue as cv
@ -345,7 +345,7 @@ class CrossFilter(object):
@type.setter
def type(self, filter_type):
if filter_type not in FILTER_TYPES.values():
if filter_type not in _FILTER_TYPES.values():
msg = 'Unable to set Filter type to "{0}" since it is not one ' \
'of the supported types'.format(type)
raise ValueError(msg)

1
openmc/filter.py
View file

@ -7,7 +7,6 @@ import numpy as np
from openmc import Mesh
from openmc.summary import Summary
from openmc.constants import *
import openmc.checkvalue as cv

198
openmc/mgxs/mgxs.py
View file

@ -190,6 +190,96 @@ class MultiGroupXS(object):
self._energy_groups = energy_groups
self._num_groups = energy_groups.num_groups
def get_all_nuclides(self):
"""Get all nuclides in the cross section's spatial domain.
Returns
-------
nuclides : list of str
A list of the string names for each nuclide in the problem domain
(e.g., ['U-235', 'U-238', 'O-16'])
Raises
------
ValueError
When this method is called before the spatial domain has been set.
"""
if self.domain is None:
raise ValueError('Unable to get all nuclides without a domain')
nuclides = self.domain.get_all_nuclides()
return nuclides.keys()
def get_nuclide_density(self, nuclide):
"""Get the atomic number density for a nuclide in the cross section's
spatial domain.
nuclide : str
A nuclide name string (e.g., 'U-235')
Returns
-------
density : Real
The atomic number density for the nuclide of interest
Raises
------
ValueError
When the density is requested for a nuclide which is not found in
the spatial domain.
"""
cv.check_type('nuclide', nuclide, basestring)
# Get list of all nuclides in the spatial domain
nuclides = self.domain.get_all_nuclides()
if nuclide not in nuclides:
msg = 'Unable to get density for nuclide "{0}" which is not in ' \
'{1} "{2}"'.format(nuclide, self.domain_type, self.domain.id)
ValueError(msg)
density = nuclides[nuclide][1]
return density
def get_nuclide_densities(self, nuclides='all'):
"""Get all atomic number densities in the cross section's spatial domain.
nuclides : Iterable of str or 'all'
A list of nuclide name strings
(e.g., ['U-235', 'U-238']; default is 'all')
Returns
-------
densities : ndarray of float
The atomic number densities corresponding to each of the nuclides
in the problem domain
Raises
------
ValueError
When this method is called before the spatial domain has been set.
"""
if self.domain is None:
raise ValueError('Unable to get nuclide densities without a domain')
# If the user requested the densities for all nuclides, get a list of
# all of the nuclide name strings
if nuclides == 'all':
nuclides =self.domain.get_all_nuclides()
# Loop over each nuclide and find and store its atomic number density
densities = np.zeros(len(nuclides), dtype=np.float)
for i, nuclide in enumerate(nuclides):
densities[i] = self.get_nuclide_density(nuclide)
return densities
@abc.abstractmethod
def create_tallies(self, scores, all_filters, keys, estimator):
"""Instantiates tallies needed to compute the multi-group cross section.
@ -243,7 +333,15 @@ class MultiGroupXS(object):
def compute_xs(self):
"""Computes multi-group cross sections using OpenMC tally arithmetic."""
return
# If a microscopic cross-section, replace CrossNuclides with originals
if self.xs_type == 'micro':
self.xs_tally._nuclides = []
nuclides = self.domain.get_all_nuclides()
for nuclide in nuclides:
self.xs_tally.add_nuclide(nuclide)
self._xs_tally._mean = np.nan_to_num(self.xs_tally.mean)
self._xs_tally._std_dev = np.nan_to_num(self.xs_tally.std_dev)
def load_from_statepoint(self, statepoint):
"""Extracts tallies in an OpenMC StatePoint with the data needed to
@ -321,8 +419,6 @@ class MultiGroupXS(object):
"""
# TODO: Multiply by densities
if self.xs_tally is None:
msg = 'Unable to get cross section since it has not been computed'
raise ValueError(msg)
@ -355,6 +451,13 @@ class MultiGroupXS(object):
# Query the multi-group cross section tally for the data
xs = self.xs_tally.get_values(filters=filters, filter_bins=filter_bins,
nuclides=nuclides, value=value)
# If user requested microscopic cross sections from an object with
# microscopic cross sections, divide by atom number densities
if self.xs_type == 'micro' and xs_type == 'micro':
densities = self.get_nuclide_densities(nuclides)
if value == 'mean' or value == 'std_dev':
xs /= densities[np.newaxis, :, np.newaxis]
return xs
def get_condensed_xs(self, coarse_groups):
@ -512,7 +615,7 @@ class MultiGroupXS(object):
return avg_xs
def print_xs(self, subdomains='all', nuclides='all'):
def print_xs(self, subdomains='all', nuclides='all', xs_type='macro'):
"""Prints a string representation for the multi-group cross section.
Parameters
@ -523,6 +626,9 @@ class MultiGroupXS(object):
nuclides : Iterable of str or 'all'
The nuclides of the cross-sections to include in the report
xs_type: {'macro' or 'micro'}
Return the macro or micro cross section in units of cm^-1 or barns
"""
if subdomains != 'all':
@ -561,11 +667,14 @@ class MultiGroupXS(object):
# Loop over all Nuclides
for nuclide in nuclides:
# Build header for cross section type based on the nuclide
if nuclide == 'total':
# Build header for nuclide type
if xs_type != 'total':
string += '{0: <16}=\t{1}\n'.format('\tNuclide', nuclide)
# Build header for cross section type
if xs_type == 'macro':
string += '{0: <16}\n'.format('\tCross Sections [cm^-1]:')
else:
string += '{0: <16}=\t{1}\n'.format('\tNuclide', nuclide)
string += '{0: <16}\n'.format('\tCross Sections [barns]:')
template = '{0: <12}Group {1} [{2: <10} - {3: <10}MeV]:\t'
@ -575,9 +684,9 @@ class MultiGroupXS(object):
bounds = self.energy_groups.get_group_bounds(group)
string += template.format('', group, bounds[0], bounds[1])
average = self.get_xs([group], [subdomain],
[nuclide], 'mean')
[nuclide], xs_type, 'mean')
rel_err = self.get_xs([group], [subdomain],
[nuclide], 'rel_err') * 100.
[nuclide], xs_type, 'rel_err') * 100
average = np.nan_to_num(average.flatten())[0]
rel_err = np.nan_to_num(rel_err.flatten())[0]
string += '{:.2e} +/- {:1.2e}%'.format(average, rel_err)
@ -646,9 +755,9 @@ class MultiGroupXS(object):
if self.xs_type == 'micro':
nuclides = self.domain.get_all_nuclides()
densities = []
for nuclide in nuclides:
densities.append(nuclides[nuclide][1])
densities = np.zeros(len(nuclides), dtype=np.float)
for i, nuclide in enumerate(nuclides):
densities[i] = nuclides[nuclide][1]
else:
nuclides = ['total']
@ -887,8 +996,7 @@ class TotalXS(MultiGroupXS):
tally arithmetic."""
self._xs_tally = self.tallies['total'] / self.tallies['flux']
self._xs_tally._mean = np.nan_to_num(self.xs_tally.mean)
self._xs_tally._std_dev = np.nan_to_num(self.xs_tally.std_dev)
super(TotalXS, self).compute_xs()
class TransportXS(MultiGroupXS):
@ -927,8 +1035,7 @@ class TransportXS(MultiGroupXS):
self._xs_tally = self.tallies['total'] - self.tallies['scatter-P1']
self._xs_tally /= self.tallies['flux']
self._xs_tally._mean = np.nan_to_num(self.xs_tally.mean)
self._xs_tally._std_dev = np.nan_to_num(self.xs_tally.std_dev)
super(TotalXS, self).compute_xs()
class AbsorptionXS(MultiGroupXS):
@ -959,8 +1066,7 @@ class AbsorptionXS(MultiGroupXS):
tally arithmetic."""
self._xs_tally = self.tallies['absorption'] / self.tallies['flux']
self._xs_tally._mean = np.nan_to_num(self.xs_tally.mean)
self._xs_tally._std_dev = np.nan_to_num(self.xs_tally.std_dev)
super(AbsorptionXS, self).compute_xs()
class CaptureXS(MultiGroupXS):
@ -992,8 +1098,7 @@ class CaptureXS(MultiGroupXS):
self._xs_tally = self.tallies['absorption'] - self.tallies['fission']
self._xs_tally /= self.tallies['flux']
self._xs_tally._mean = np.nan_to_num(self.xs_tally.mean)
self._xs_tally._std_dev = np.nan_to_num(self.xs_tally.std_dev)
super(CaptureXS, self).compute_xs()
class FissionXS(MultiGroupXS):
@ -1024,8 +1129,7 @@ class FissionXS(MultiGroupXS):
tally arithmetic."""
self._xs_tally = self.tallies['fission'] / self.tallies['flux']
self._xs_tally._mean = np.nan_to_num(self.xs_tally.mean)
self._xs_tally._std_dev = np.nan_to_num(self.xs_tally.std_dev)
super(FissionXS, self).compute_xs()
class NuFissionXS(MultiGroupXS):
@ -1056,8 +1160,7 @@ class NuFissionXS(MultiGroupXS):
tally arithmetic."""
self._xs_tally = self.tallies['nu-fission'] / self.tallies['flux']
self._xs_tally._mean = np.nan_to_num(self.xs_tally.mean)
self._xs_tally._std_dev = np.nan_to_num(self.xs_tally.std_dev)
super(NuFissionXS, self).compute_xs()
class ScatterXS(MultiGroupXS):
@ -1088,8 +1191,7 @@ class ScatterXS(MultiGroupXS):
OpenMC tally arithmetic."""
self._xs_tally = self.tallies['scatter'] / self.tallies['flux']
self._xs_tally._mean = np.nan_to_num(self.xs_tally.mean)
self._xs_tally._std_dev = np.nan_to_num(self.xs_tally.std_dev)
super(ScatterXS, self).compute_xs()
class NuScatterXS(MultiGroupXS):
@ -1120,8 +1222,7 @@ class NuScatterXS(MultiGroupXS):
tally arithmetic."""
self._xs_tally = self.tallies['nu-scatter'] / self.tallies['flux']
self._xs_tally._mean = np.nan_to_num(self.xs_tally.mean)
self._xs_tally._std_dev = np.nan_to_num(self.xs_tally.std_dev)
super(NuScatterXS, self).compute_xs()
class ScatterMatrixXS(MultiGroupXS):
@ -1172,8 +1273,7 @@ class ScatterMatrixXS(MultiGroupXS):
rxn_tally = self.tallies['scatter']
self._xs_tally = rxn_tally / self.tallies['flux']
self._xs_tally._mean = np.nan_to_num(self.xs_tally.mean)
self._xs_tally._std_dev = np.nan_to_num(self.xs_tally.std_dev)
super(ScatterMatrixXS, self).compute_xs()
def get_xs(self, in_groups='all', out_groups='all', subdomains='all',
nuclides='all', xs_type='macro', value='mean'):
@ -1218,8 +1318,6 @@ class ScatterMatrixXS(MultiGroupXS):
"""
# TODO: Deal with xs_type and micros
if self.xs_tally is None:
msg = 'Unable to get cross section since it has not been computed'
raise ValueError(msg)
@ -1260,9 +1358,17 @@ class ScatterMatrixXS(MultiGroupXS):
xs = self.xs_tally.get_values(filters=filters, filter_bins=filter_bins,
nuclides=nuclides, value=value)
xs = np.nan_to_num(xs)
# If user requested microscopic cross sections from an object with
# microscopic cross sections, divide by atom number densities
if self.xs_type == 'micro' and xs_type == 'micro':
densities = self.get_nuclide_densities(nuclides)
if value == 'mean' or value == 'std_dev':
xs /= densities[np.newaxis, :, np.newaxis]
return xs
def print_xs(self, subdomains='all', nuclides='all'):
def print_xs(self, subdomains='all', nuclides='all', xs_type='macro'):
"""Prints a string representation for the multi-group cross section.
Parameters
@ -1273,6 +1379,9 @@ class ScatterMatrixXS(MultiGroupXS):
nuclides : Iterable of str or 'all'
The nuclides of the cross-sections to include in the report
xs_type: {'macro' or 'micro'}
Return the macro or micro cross section in units of cm^-1 or barns
"""
if subdomains != 'all':
@ -1320,11 +1429,14 @@ class ScatterMatrixXS(MultiGroupXS):
# Loop over all Nuclides
for nuclide in nuclides:
# Build header for cross section type based on the nuclide
if nuclide == 'total':
# Build header for nuclide type
if xs_type != 'total':
string += '{0: <16}=\t{1}\n'.format('\tNuclide', nuclide)
# Build header for cross section type
if xs_type == 'macro':
string += '{0: <16}\n'.format('\tCross Sections [cm^-1]:')
else:
string += '{0: <16}=\t{1}\n'.format('\tNuclide', nuclide)
string += '{0: <16}\n'.format('\tCross Sections [barns]:')
template = '{0: <12}Group {1} -> Group {2}:\t\t'
@ -1334,11 +1446,11 @@ class ScatterMatrixXS(MultiGroupXS):
for out_group in range(1, self.num_groups+1):
string += template.format('', in_group, out_group)
average = \
self.get_xs([in_group], [out_group],
[subdomain], [nuclide], 'mean')
self.get_xs([in_group], [out_group], [subdomain],
[nuclide], xs_type, 'mean')
rel_err = \
self.get_xs([in_group], [out_group],
[subdomain], [nuclide],'rel_err') * 100
self.get_xs([in_group], [out_group], [subdomain],
[nuclide], xs_type, 'rel_err') * 100
average = np.nan_to_num(average.flatten())[0]
rel_err = np.nan_to_num(rel_err.flatten())[0]
string += '{:1.2e} +/- {:1.2e}%'.format(average, rel_err)
@ -1398,8 +1510,7 @@ class NuScatterMatrixXS(ScatterMatrixXS):
rxn_tally = self.tallies['nu-scatter']
self._xs_tally = rxn_tally / self.tallies['flux']
self._xs_tally._mean = np.nan_to_num(self.xs_tally.mean)
self._xs_tally._std_dev = np.nan_to_num(self.xs_tally.std_dev)
super(ScatterMatrixXS, self).compute_xs()
class Chi(MultiGroupXS):
@ -1434,8 +1545,7 @@ class Chi(MultiGroupXS):
nu_fission_out = self.tallies['nu-fission-out']
nu_fission_in.remove_filter(nu_fission_in.filters[-1])
self._xs_tally = nu_fission_out / nu_fission_in
self._xs_tally._mean = np.nan_to_num(self.xs_tally.mean)
self._xs_tally._std_dev = np.nan_to_num(self.xs_tally.std_dev)
super(Chi, self).compute_xs()
def get_xs(self, groups='all', subdomains='all',
nuclides='all', xs_type='macro', value='mean'):

5
openmc/tallies.py
View file

@ -1,4 +1,4 @@
from collections import Iterable, defaultdict
from collections import Iterable
import copy
import os
import pickle
@ -9,9 +9,8 @@ import sys
import numpy as np
from openmc import Mesh, Filter, Trigger, Nuclide, FILTER_TYPES
from openmc import Mesh, Filter, Trigger, Nuclide
from openmc.cross import CrossScore, CrossNuclide, CrossFilter
from openmc.summary import Summary
import openmc.checkvalue as cv
from openmc.clean_xml import *