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incorporated MT=5, simplified yield usage in the plotter

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This commit is contained in:
Adam Nelson 2016-11-18 19:55:24 -05:00
parent 1a62b59dc3
commit b4c556db82
3 changed files with 36 additions and 39 deletions
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3
openmc/data/library.py
View file

@ -1,5 +1,6 @@
import os
import xml.etree.ElementTree as ET
from six import string_types
import h5py
@ -124,7 +125,7 @@ class DataLibrary(EqualityMixin):
raise ValueError("Either path or OPENMC_CROSS_SECTIONS "
"environmental variable must be set")
check_type('path', path, str)
check_type('path', path, string_types)
tree = ET.parse(path)
root = tree.getroot()

7
openmc/material.py
View file

@ -606,15 +606,10 @@ class Material(object):
return nuclides
def get_nuclide_atom_densities(self, cross_sections=None):
def get_nuclide_atom_densities(self):
"""Returns all nuclides in the material and their atomic densities in
units of atom/b-cm
Parameters
----------
cross_sections : str, optional
Location of cross_sections.xml file. Default is None.
Returns
-------
nuclides : dict

65
openmc/plotter.py
View file

@ -1,4 +1,5 @@
from numbers import Integral, Real
from six import string_types
import numpy as np
@ -15,7 +16,7 @@ UNITY_MT = -1
XI_MT = -2
# MTs to combine to generate associated plot_types
_INELASTIC = [mt for mt in openmc.data.SUM_RULES[3] if mt not in [5, 27]]
_INELASTIC = [mt for mt in openmc.data.SUM_RULES[3] if mt != 27]
PLOT_TYPES_MT = {'total': openmc.data.SUM_RULES[1],
'scatter': [2] + _INELASTIC,
'elastic': [2],
@ -41,19 +42,6 @@ PLOT_TYPES_OP = {'total': (np.add,),
(np.add,) * (len(PLOT_TYPES_MT['slowing-down power']) - 2) + (np.multiply,),
'damage': ()}
# Whether or not to multiply the reaction by the yield as well
PLOT_TYPES_YIELD = {'total': (False, False),
'scatter': (False,) * len(PLOT_TYPES_MT['scatter']),
'elastic': (False,),
'inelastic': (False,) * len(PLOT_TYPES_MT['inelastic']),
'fission': (False,), 'absorption': (False,),
'capture': (False,), 'nu-fission': (True,),
'nu-scatter': (True,) * len(PLOT_TYPES_MT['nu-scatter']),
'unity': (False,),
'slowing-down power':
(True,) * len(PLOT_TYPES_MT['slowing-down power']),
'damage': (False,)}
# Types of plots to plot linearly in y
PLOT_TYPES_LINEAR = {'nu-fission / fission', 'nu-scatter / scatter',
'nu-fission / absorption', 'fission / absorption'}
@ -230,7 +218,7 @@ def calculate_xs(this, types, temperature=294., sab_name=None,
# Check types
cv.check_type('temperature', temperature, Real)
if sab_name:
cv.check_type('sab_name', sab_name, str)
cv.check_type('sab_name', sab_name, string_types)
if enrichment:
cv.check_type('enrichment', enrichment, Real)
@ -370,15 +358,18 @@ def _calculate_xs_nuclide(this, types, temperature=294., sab_name=None,
for line in types:
if line in PLOT_TYPES:
mts.append(PLOT_TYPES_MT[line])
yields.append(PLOT_TYPES_YIELD[line])
if line.startswith('nu'):
yields.append(True)
else:
yields.append(False)
ops.append(PLOT_TYPES_OP[line])
else:
# Not a built-in type, we have to parse it ourselves
cv.check_type('MT in types', line, Integral)
cv.check_greater_than('MT in types', line, 0)
mts.append((line,))
yields.append((False,))
ops.append(())
yields.append(False)
# Load the library
library = openmc.data.DataLibrary.from_xml(cross_sections)
@ -456,7 +447,7 @@ def _calculate_xs_nuclide(this, types, temperature=294., sab_name=None,
# Get the reaction xs data from the nuclide
funcs = []
op = ops[i]
for mt, yield_check in zip(mt_set, yields[i]):
for mt in mt_set:
if mt == 2:
if sab_name:
# Then we need to do a piece-wise function of
@ -468,28 +459,38 @@ def _calculate_xs_nuclide(this, types, temperature=294., sab_name=None,
funcs.append(pw_funcs)
else:
funcs.append(nuc[mt].xs[nucT])
elif mt == 5 and mt in nuc:
# Only consider the (n,misc) products if neutrons are
# included in the outgoing channel since (n,misc) is only
# explicitly needed for scatter cross sections.
for prod in nuc[mt].products:
if prod.particle == 'neutron' and \
prod.emission_mode in ('total', 'prompt'):
if yields[i]:
func = openmc.data.Combination(
[nuc[mt].xs[nucT], prod.yield_],
[np.multiply])
else:
func = nuc[mt].xs[nucT]
funcs.append(func)
break
else:
funcs.append(lambda x: 0.)
elif mt in nuc:
if yield_check:
if yields[i]:
for prod in nuc[mt].products:
if prod.particle == 'neutron' and \
prod.emission_mode == 'total':
prod.emission_mode in ('total', 'prompt'):
func = openmc.data.Combination(
[nuc[mt].xs[nucT], prod.yield_],
[np.multiply])
funcs.append(func)
break
else:
for prod in nuc[mt].products:
if prod.particle == 'neutron' and \
prod.emission_mode == 'prompt':
func = openmc.data.Combination(
[nuc[mt].xs[nucT],
prod.yield_], [np.multiply])
funcs.append(func)
break
else:
# Assume the yield is 1
funcs.append(nuc[mt].xs[nucT])
# Assume the yield is 1
funcs.append(nuc[mt].xs[nucT])
else:
funcs.append(nuc[mt].xs[nucT])
elif mt == UNITY_MT:
@ -545,7 +546,7 @@ def _calculate_xs_material(this, types, temperature=294., cross_sections=None):
library = openmc.data.DataLibrary.from_xml(cross_sections)
# Expand elements in to nuclides with atomic densities
nuclides = this.get_nuclide_atom_densities(cross_sections)
nuclides = this.get_nuclide_atom_densities()
# For ease of processing split out nuc and nuc_density
nuc_densities = [nuclide[1][1] for nuclide in nuclides.items()]