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Merge remote-tracking branch 'upstream/develop' into 2d-mesh-tally

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This commit is contained in:
Sam Shaner 2016-09-12 09:50:17 -04:00
commit 5affc137e1
180 changed files with 4667 additions and 4932 deletions
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3
examples/python/basic/build-xml.py
View file

@ -25,7 +25,7 @@ moderator = openmc.Material(material_id=41, name='moderator')
moderator.set_density('g/cc', 1.0)
moderator.add_nuclide(h1, 2.)
moderator.add_nuclide(o16, 1.)
moderator.add_s_alpha_beta('c_H_in_H2O', '71t')
moderator.add_s_alpha_beta('c_H_in_H2O')
fuel = openmc.Material(material_id=40, name='fuel')
fuel.set_density('g/cc', 4.5)
@ -33,7 +33,6 @@ fuel.add_nuclide(u235, 1.)
# Instantiate a Materials collection and export to XML
materials_file = openmc.Materials([moderator, fuel])
materials_file.default_xs = '71c'
materials_file.export_to_xml()

3
examples/python/boxes/build-xml.py
View file

@ -34,11 +34,10 @@ moderator = openmc.Material(material_id=3, name='moderator')
moderator.set_density('g/cc', 1.0)
moderator.add_nuclide(h1, 2.)
moderator.add_nuclide(o16, 1.)
moderator.add_s_alpha_beta('c_H_in_H2O', '71t')
moderator.add_s_alpha_beta('c_H_in_H2O')
# Instantiate a Materials collection and export to XML
materials_file = openmc.Materials([fuel1, fuel2, moderator])
materials_file.default_xs = '71c'
materials_file.export_to_xml()

3
examples/python/lattice/hexagonal/build-xml.py
View file

@ -29,7 +29,7 @@ moderator = openmc.Material(material_id=2, name='moderator')
moderator.set_density('g/cc', 1.0)
moderator.add_nuclide(h1, 2.)
moderator.add_nuclide(o16, 1.)
moderator.add_s_alpha_beta('c_H_in_H2O', '71t')
moderator.add_s_alpha_beta('c_H_in_H2O')
iron = openmc.Material(material_id=3, name='iron')
iron.set_density('g/cc', 7.9)
@ -37,7 +37,6 @@ iron.add_nuclide(fe56, 1.)
# Instantiate a Materials collection and export to XML
materials_file = openmc.Materials([moderator, fuel, iron])
materials_file.default_xs = '71c'
materials_file.export_to_xml()

3
examples/python/lattice/nested/build-xml.py
View file

@ -28,11 +28,10 @@ moderator = openmc.Material(material_id=2, name='moderator')
moderator.set_density('g/cc', 1.0)
moderator.add_nuclide(h1, 2.)
moderator.add_nuclide(o16, 1.)
moderator.add_s_alpha_beta('c_H_in_H2O', '71t')
moderator.add_s_alpha_beta('c_H_in_H2O')
# Instantiate a Materials collection and export to XML
materials_file = openmc.Materials((moderator, fuel))
materials_file.default_xs = '71c'
materials_file.export_to_xml()

3
examples/python/lattice/simple/build-xml.py
View file

@ -28,11 +28,10 @@ moderator = openmc.Material(material_id=2, name='moderator')
moderator.set_density('g/cc', 1.0)
moderator.add_nuclide(h1, 2.)
moderator.add_nuclide(o16, 1.)
moderator.add_s_alpha_beta('c_H_in_H2O', '71t')
moderator.add_s_alpha_beta('c_H_in_H2O')
# Instantiate a Materials collection and export to XML
materials_file = openmc.Materials([moderator, fuel])
materials_file.default_xs = '71c'
materials_file.export_to_xml()

3
examples/python/pincell/build-xml.py
View file

@ -98,11 +98,10 @@ borated_water.add_nuclide(h1, 4.9457e-2)
borated_water.add_nuclide(h2, 7.4196e-6)
borated_water.add_nuclide(o16, 2.4672e-2)
borated_water.add_nuclide(o17, 6.0099e-5)
borated_water.add_s_alpha_beta('c_H_in_H2O', '71t')
borated_water.add_s_alpha_beta('c_H_in_H2O')
# Instantiate a Materials collection and export to XML
materials_file = openmc.Materials([uo2, helium, zircaloy, borated_water])
materials_file.default_xs = '71c'
materials_file.export_to_xml()

9
examples/python/pincell_multigroup/build-xml.py
View file

@ -19,7 +19,7 @@ groups = openmc.mgxs.EnergyGroups(group_edges=[1E-11, 0.0635E-6, 10.0E-6,
1.0E-4, 1.0E-3, 0.5, 1.0, 20.0])
# Instantiate the 7-group (C5G7) cross section data
uo2_xsdata = openmc.XSdata('UO2.300K', groups)
uo2_xsdata = openmc.XSdata('UO2', groups)
uo2_xsdata.order = 0
uo2_xsdata.total = [0.1779492, 0.3298048, 0.4803882, 0.5543674,
0.3118013, 0.3951678, 0.5644058]
@ -41,7 +41,7 @@ uo2_xsdata.nu_fission = [2.005998E-02, 2.027303E-03, 1.570599E-02,
uo2_xsdata.chi = [5.8791E-01, 4.1176E-01, 3.3906E-04, 1.1761E-07,
0.0000E+00, 0.0000E+00, 0.0000E+00]
h2o_xsdata = openmc.XSdata('LWTR.300K', groups)
h2o_xsdata = openmc.XSdata('LWTR', groups)
h2o_xsdata.order = 0
h2o_xsdata.total = [0.15920605, 0.412969593, 0.59030986, 0.58435,
0.718, 1.2544497, 2.650379]
@ -66,8 +66,8 @@ mg_cross_sections_file.export_to_xml()
###############################################################################
# Instantiate some Macroscopic Data
uo2_data = openmc.Macroscopic('UO2', '300K')
h2o_data = openmc.Macroscopic('LWTR', '300K')
uo2_data = openmc.Macroscopic('UO2')
h2o_data = openmc.Macroscopic('LWTR')
# Instantiate some Materials and register the appropriate Macroscopic objects
uo2 = openmc.Material(material_id=1, name='UO2 fuel')
@ -80,7 +80,6 @@ water.add_macroscopic(h2o_data)
# Instantiate a Materials collection and export to XML
materials_file = openmc.Materials([uo2, water])
materials_file.default_xs = '300K'
materials_file.export_to_xml()

1
examples/python/reflective/build-xml.py
View file

@ -25,7 +25,6 @@ fuel.add_nuclide(u235, 1.)
# Instantiate a Materials collection and export to XML
materials_file = openmc.Materials([fuel])
materials_file.default_xs = '71c'
materials_file.export_to_xml()

4
examples/xml/basic/materials.xml
View file

@ -1,8 +1,6 @@
<?xml version="1.0"?>
<materials>
<default_xs>71c</default_xs>
<material id="40">
<density value="4.5" units="g/cc" />
<nuclide name="U235" ao="1.0" />
@ -12,7 +10,7 @@
<density value="1.0" units="g/cc" />
<nuclide name="H1" ao="2.0" />
<nuclide name="O16" ao="1.0" />
<sab name="c_H_in_H2O" xs="71t" />
<sab name="c_H_in_H2O"/>
</material>
</materials>

4
examples/xml/boxes/materials.xml
View file

@ -1,8 +1,6 @@
<?xml version="1.0"?>
<materials>
<default_xs>71c</default_xs>
<material id="1">
<density value="4.5" units="g/cc" />
<nuclide name="U235" ao="1.0" />
@ -17,7 +15,7 @@
<density value="1.0" units="g/cc" />
<nuclide name="O16" ao="1.0" />
<nuclide name="H1" ao="2.0" />
<sab name="c_H_in_H2O" xs="71t" />
<sab name="c_H_in_H2O" />
</material>
</materials>

4
examples/xml/lattice/nested/materials.xml
View file

@ -1,8 +1,6 @@
<?xml version="1.0"?>
<materials>
<default_xs>71c</default_xs>
<!-- Definition of materials -->
<material id="1">
<density value="4.5" units="g/cc" />
@ -13,7 +11,7 @@
<density value="1.0" units="g/cc" />
<nuclide name="H1" ao="2.0" />
<nuclide name="O16" ao="1.0" />
<sab name="c_H_in_H2O" xs="71t" />
<sab name="c_H_in_H2O" />
</material>
</materials>

4
examples/xml/lattice/simple/materials.xml
View file

@ -1,8 +1,6 @@
<?xml version="1.0"?>
<materials>
<default_xs>71c</default_xs>
<!-- Definition of materials -->
<material id="1">
<density value="4.5" units="g/cc" />
@ -13,7 +11,7 @@
<density value="1.0" units="g/cc" />
<nuclide name="H1" ao="2.0" />
<nuclide name="O16" ao="1.0" />
<sab name="c_H_in_H2O" xs="71t" />
<sab name="c_H_in_H2O" />
</material>
</materials>

5
examples/xml/pincell/materials.xml
View file

@ -1,9 +1,6 @@
<?xml version="1.0"?>
<materials>
<!-- By default, use 300K cross sections -->
<default_xs>71c</default_xs>
<!--
Since O-18 is not present in ENDF/B-VII, it was necessary to combine the
atom densities for O-17 and O-18 in any materials containing Oxygen.
@ -64,7 +61,7 @@
<nuclide name="H2" ao="7.4196e-06" />
<nuclide name="O16" ao="2.4672e-02" />
<nuclide name="O17" ao="6.0099e-05" />
<sab name="c_H_in_H2O" xs="71t" />
<sab name="c_H_in_H2O" />
</material>
</materials>

3
examples/xml/pincell_multigroup/materials.xml
View file

@ -1,8 +1,5 @@
<?xml version="1.0"?>
<materials>
<!-- Set default xs set to use 300K data -->
<default_xs>300K</default_xs>
<!-- UO2 -->
<material id="1">
<density units="macro" value="1.0" />

32
examples/xml/pincell_multigroup/mg_cross_sections.xml
View file

@ -11,8 +11,8 @@
-->
<xsdata>
<!-- Meta data for this data -->
<name>UO2.300K</name>
<alias>UO2.300K</alias>
<name>UO2</name>
<alias>UO2</alias>
<kT> 2.53E-8 </kT> <!-- in MeV -->
<order>0</order>
<fissionable>true</fissionable>
@ -67,8 +67,8 @@
<xsdata>
<!-- Meta data for this data -->
<name>MOX1.300K</name>
<alias>MOX1.300K</alias>
<name>MOX1</name>
<alias>MOX1</alias>
<kT> 2.53E-8 </kT> <!-- in MeV -->
<order>0</order>
<fissionable>true</fissionable>
@ -124,8 +124,8 @@
<xsdata>
<!-- Meta data for this data -->
<name>MOX2.300K</name>
<alias>MOX2.300K</alias>
<name>MOX2</name>
<alias>MOX2</alias>
<kT> 2.53E-8 </kT> <!-- in MeV -->
<order>0</order>
<fissionable>true</fissionable>
@ -180,8 +180,8 @@
<xsdata>
<!-- Meta data for this data -->
<name>MOX3.300K</name>
<alias>MOX3.300K</alias>
<name>MOX3</name>
<alias>MOX3</alias>
<kT> 2.53E-8 </kT> <!-- in MeV -->
<order>0</order>
<fissionable>true</fissionable>
@ -236,8 +236,8 @@
<xsdata>
<!-- Meta data for this data -->
<name>FC.300K</name>
<alias>FC.300K</alias>
<name>FC</name>
<alias>FC</alias>
<kT> 2.53E-8 </kT> <!-- in MeV -->
<order>0</order>
<fissionable>true</fissionable>
@ -286,8 +286,8 @@
<xsdata>
<!-- Meta data for this data -->
<name>GT.300K</name>
<alias>GT.300K</alias>
<name>GT</name>
<alias>GT</alias>
<kT> 2.53E-8 </kT> <!-- in MeV -->
<order>0</order>
<fissionable>false</fissionable>
@ -318,8 +318,8 @@
<xsdata>
<!-- Meta data for this data -->
<name>LWTR.300K</name>
<alias>LWTR.300K</alias>
<name>LWTR</name>
<alias>LWTR</alias>
<kT> 2.53E-8 </kT> <!-- in MeV -->
<order>0</order>
<fissionable>false</fissionable>
@ -351,8 +351,8 @@
<xsdata>
<!-- Meta data for this data -->
<name>CR.300K</name>
<alias>CR.300K</alias>
<name>CR</name>
<alias>CR</alias>
<kT> 2.53E-8 </kT> <!-- in MeV -->
<order>0</order>
<fissionable>false</fissionable>

2
examples/xml/reflective/materials.xml
View file

@ -1,8 +1,6 @@
<?xml version="1.0"?>
<materials>
<default_xs>71c</default_xs>
<material id="1">
<density value="4.5" units="g/cc" />
<nuclide name="U235" ao="1.0" />