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updated tests and updated setting of nuclides and elements in example files

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This commit is contained in:
samuelshaner 2016-12-15 21:06:12 +00:00
parent ff3d034275
commit 1a7630083b
18 changed files with 114 additions and 201 deletions
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11
examples/python/basic/build-xml.py
View file

@ -15,21 +15,16 @@ particles = 10000
# Exporting to OpenMC materials.xml file
###############################################################################
# Instantiate some Nuclides
h1 = openmc.Nuclide('H1')
o16 = openmc.Nuclide('O16')
u235 = openmc.Nuclide('U235')
# Instantiate some Materials and register the appropriate Nuclides
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_element('H', 2.)
moderator.add_element('O', 1.)
moderator.add_s_alpha_beta('c_H_in_H2O')
fuel = openmc.Material(material_id=40, name='fuel')
fuel.set_density('g/cc', 4.5)
fuel.add_nuclide(u235, 1.)
fuel.add_nuclide('U235', 1.)
# Instantiate a Materials collection and export to XML
materials_file = openmc.Materials([moderator, fuel])

14
examples/python/boxes/build-xml.py
View file

@ -15,25 +15,19 @@ particles = 10000
# Exporting to OpenMC materials.xml File
###############################################################################
# Instantiate some Nuclides
h1 = openmc.Nuclide('H1')
o16 = openmc.Nuclide('O16')
u235 = openmc.Nuclide('U235')
u238 = openmc.Nuclide('U238')
# Instantiate some Materials and register the appropriate Nuclides
fuel1 = openmc.Material(material_id=1, name='fuel')
fuel1.set_density('g/cc', 4.5)
fuel1.add_nuclide(u235, 1.)
fuel1.add_nuclide('U235', 1.)
fuel2 = openmc.Material(material_id=2, name='depleted fuel')
fuel2.set_density('g/cc', 4.5)
fuel2.add_nuclide(u238, 1.)
fuel2.add_nuclide('U238', 1.)
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_element('H', 2.)
moderator.add_element('O', 1.)
moderator.add_s_alpha_beta('c_H_in_H2O')
# Instantiate a Materials collection and export to XML

14
examples/python/lattice/hexagonal/build-xml.py
View file

@ -14,26 +14,20 @@ particles = 10000
# Exporting to OpenMC materials.xml File
###############################################################################
# Instantiate some Nuclides
h1 = openmc.Nuclide('H1')
o16 = openmc.Nuclide('O16')
u235 = openmc.Nuclide('U235')
fe56 = openmc.Nuclide('Fe56')
# Instantiate some Materials and register the appropriate Nuclides
fuel = openmc.Material(material_id=1, name='fuel')
fuel.set_density('g/cc', 4.5)
fuel.add_nuclide(u235, 1.)
fuel.add_nuclide('U235', 1.)
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_element('H', 2.)
moderator.add_element('O', 1.)
moderator.add_s_alpha_beta('c_H_in_H2O')
iron = openmc.Material(material_id=3, name='iron')
iron.set_density('g/cc', 7.9)
iron.add_nuclide(fe56, 1.)
iron.add_element('Fe', 1.)
# Instantiate a Materials collection and export to XML
materials_file = openmc.Materials([moderator, fuel, iron])

11
examples/python/lattice/nested/build-xml.py
View file

@ -14,20 +14,15 @@ particles = 10000
# Exporting to OpenMC materials.xml file
###############################################################################
# Instantiate some Nuclides
h1 = openmc.Nuclide('H1')
o16 = openmc.Nuclide('O16')
u235 = openmc.Nuclide('U235')
# Instantiate some Materials and register the appropriate Nuclides
fuel = openmc.Material(material_id=1, name='fuel')
fuel.set_density('g/cc', 4.5)
fuel.add_nuclide(u235, 1.)
fuel.add_nuclide('U235', 1.)
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_element('H', 2.)
moderator.add_element('O', 1.)
moderator.add_s_alpha_beta('c_H_in_H2O')
# Instantiate a Materials collection and export to XML

11
examples/python/lattice/simple/build-xml.py
View file

@ -14,20 +14,15 @@ particles = 10000
# Exporting to OpenMC materials.xml file
###############################################################################
# Instantiate some Nuclides
h1 = openmc.Nuclide('H1')
o16 = openmc.Nuclide('O16')
u235 = openmc.Nuclide('U235')
# Instantiate some Materials and register the appropriate Nuclides
fuel = openmc.Material(material_id=1, name='fuel')
fuel.set_density('g/cc', 4.5)
fuel.add_nuclide(u235, 1.)
fuel.add_nuclide('U235', 1.)
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_element('H', 2.)
moderator.add_element('O', 1.)
moderator.add_s_alpha_beta('c_H_in_H2O')
# Instantiate a Materials collection and export to XML

79
examples/python/pincell/build-xml.py
View file

@ -14,86 +14,29 @@ particles = 1000
# Exporting to OpenMC materials.xml file
###############################################################################
# Instantiate some Nuclides
h1 = openmc.Nuclide('H1')
h2 = openmc.Nuclide('H2')
he4 = openmc.Nuclide('He4')
b10 = openmc.Nuclide('B10')
b11 = openmc.Nuclide('B11')
o16 = openmc.Nuclide('O16')
o17 = openmc.Nuclide('O17')
cr50 = openmc.Nuclide('Cr50')
cr52 = openmc.Nuclide('Cr52')
cr53 = openmc.Nuclide('Cr53')
cr54 = openmc.Nuclide('Cr54')
fe54 = openmc.Nuclide('Fe54')
fe56 = openmc.Nuclide('Fe56')
fe57 = openmc.Nuclide('Fe57')
fe58 = openmc.Nuclide('Fe58')
zr90 = openmc.Nuclide('Zr90')
zr91 = openmc.Nuclide('Zr91')
zr92 = openmc.Nuclide('Zr92')
zr94 = openmc.Nuclide('Zr94')
zr96 = openmc.Nuclide('Zr96')
sn112 = openmc.Nuclide('Sn112')
sn114 = openmc.Nuclide('Sn114')
sn115 = openmc.Nuclide('Sn115')
sn116 = openmc.Nuclide('Sn116')
sn117 = openmc.Nuclide('Sn117')
sn118 = openmc.Nuclide('Sn118')
sn119 = openmc.Nuclide('Sn119')
sn120 = openmc.Nuclide('Sn120')
sn122 = openmc.Nuclide('Sn122')
sn124 = openmc.Nuclide('Sn124')
u = openmc.Element('U')
o = openmc.Element('O')
# Instantiate some Materials and register the appropriate Nuclides
uo2 = openmc.Material(material_id=1, name='UO2 fuel at 2.4% wt enrichment')
uo2.set_density('g/cm3', 10.29769)
uo2.add_element(u, 1., enrichment=2.4)
uo2.add_element(o, 2.)
uo2.add_element('U', 1., enrichment=2.4)
uo2.add_element('O', 2.)
helium = openmc.Material(material_id=2, name='Helium for gap')
helium.set_density('g/cm3', 0.001598)
helium.add_nuclide(he4, 2.4044e-4)
helium.add_element('He', 2.4044e-4)
zircaloy = openmc.Material(material_id=3, name='Zircaloy 4')
zircaloy.set_density('g/cm3', 6.55)
zircaloy.add_nuclide(o16, 3.0743e-4)
zircaloy.add_nuclide(o17, 7.4887e-7)
zircaloy.add_nuclide(cr50, 3.2962e-6)
zircaloy.add_nuclide(cr52, 6.3564e-5)
zircaloy.add_nuclide(cr53, 7.2076e-6)
zircaloy.add_nuclide(cr54, 1.7941e-6)
zircaloy.add_nuclide(fe54, 8.6699e-6)
zircaloy.add_nuclide(fe56, 1.3610e-4)
zircaloy.add_nuclide(fe57, 3.1431e-6)
zircaloy.add_nuclide(fe58, 4.1829e-7)
zircaloy.add_nuclide(zr90, 2.1827e-2)
zircaloy.add_nuclide(zr91, 4.7600e-3)
zircaloy.add_nuclide(zr92, 7.2758e-3)
zircaloy.add_nuclide(zr94, 7.3734e-3)
zircaloy.add_nuclide(zr96, 1.1879e-3)
zircaloy.add_nuclide(sn112, 4.6735e-6)
zircaloy.add_nuclide(sn114, 3.1799e-6)
zircaloy.add_nuclide(sn115, 1.6381e-6)
zircaloy.add_nuclide(sn116, 7.0055e-5)
zircaloy.add_nuclide(sn117, 3.7003e-5)
zircaloy.add_nuclide(sn118, 1.1669e-4)
zircaloy.add_nuclide(sn119, 4.1387e-5)
zircaloy.add_nuclide(sn120, 1.5697e-4)
zircaloy.add_nuclide(sn122, 2.2308e-5)
zircaloy.add_nuclide(sn124, 2.7897e-5)
zircaloy.add_element('Sn', 0.014 , 'wo')
zircaloy.add_element('Fe', 0.00165, 'wo')
zircaloy.add_element('Cr', 0.001 , 'wo')
zircaloy.add_element('Zr', 0.98335, 'wo')
borated_water = openmc.Material(material_id=4, name='Borated water at 975 ppm')
borated_water = openmc.Material(material_id=4, name='Borated water')
borated_water.set_density('g/cm3', 0.740582)
borated_water.add_nuclide(b10, 8.0042e-6)
borated_water.add_nuclide(b11, 3.2218e-5)
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_element('B', 4.0e-5)
borated_water.add_element('H', 5.0e-2)
borated_water.add_element('O', 2.4e-2)
borated_water.add_s_alpha_beta('c_H_in_H2O')
# Instantiate a Materials collection and export to XML

5
examples/python/reflective/build-xml.py
View file

@ -15,13 +15,10 @@ particles = 10000
# Exporting to OpenMC materials.xml file
###############################################################################
# Instantiate a Nuclides
u235 = openmc.Nuclide('U235')
# Instantiate a Material and register the Nuclide
fuel = openmc.Material(material_id=1, name='fuel')
fuel.set_density('g/cc', 4.5)
fuel.add_nuclide(u235, 1.)
fuel.add_nuclide('U235', 1.)
# Instantiate a Materials collection and export to XML
materials_file = openmc.Materials([fuel])

2
openmc/settings.py
View file

@ -830,7 +830,7 @@ class Settings(object):
def _create_keff_trigger_subelement(self, run_mode_element):
if self._keff_trigger is not None:
element = ET.SubElement(run_mode_subelement, "keff_trigger")
element = ET.SubElement(run_mode_element, "keff_trigger")
for key in self._keff_trigger:
subelement = ET.SubElement(element, key)

12
tests/test_mgxs_library_condense/inputs_true.dat
View file

@ -396,7 +396,7 @@
<filter bins="0.0 0.625 20000000.0" type="energy" />
<nuclides>total</nuclides>
<scores>delayed-nu-fission</scores>
<estimator>analog</estimator>
<estimator>tracklength</estimator>
</tally>
<tally id="10048">
<filter bins="10000" type="material" />
@ -404,7 +404,7 @@
<filter bins="0.0 0.625 20000000.0" type="energy" />
<nuclides>total</nuclides>
<scores>decay-rate</scores>
<estimator>analog</estimator>
<estimator>tracklength</estimator>
</tally>
<tally id="10049">
<filter bins="10000" type="material" />
@ -767,7 +767,7 @@
<filter bins="0.0 0.625 20000000.0" type="energy" />
<nuclides>total</nuclides>
<scores>delayed-nu-fission</scores>
<estimator>analog</estimator>
<estimator>tracklength</estimator>
</tally>
<tally id="10099">
<filter bins="10001" type="material" />
@ -775,7 +775,7 @@
<filter bins="0.0 0.625 20000000.0" type="energy" />
<nuclides>total</nuclides>
<scores>decay-rate</scores>
<estimator>analog</estimator>
<estimator>tracklength</estimator>
</tally>
<tally id="10100">
<filter bins="10001" type="material" />
@ -1138,7 +1138,7 @@
<filter bins="0.0 0.625 20000000.0" type="energy" />
<nuclides>total</nuclides>
<scores>delayed-nu-fission</scores>
<estimator>analog</estimator>
<estimator>tracklength</estimator>
</tally>
<tally id="10150">
<filter bins="10002" type="material" />
@ -1146,7 +1146,7 @@
<filter bins="0.0 0.625 20000000.0" type="energy" />
<nuclides>total</nuclides>
<scores>decay-rate</scores>
<estimator>analog</estimator>
<estimator>tracklength</estimator>
</tally>
<tally id="10151">
<filter bins="10002" type="material" />

12
tests/test_mgxs_library_condense/results_true.dat
View file

@ -64,12 +64,12 @@
4 10000 5 1 total 0.001231 0.000105
5 10000 6 1 total 0.000512 0.000044
material delayedgroup group in nuclide mean std. dev.
0 10000 1 1 total 0.000000 0.000000
1 10000 2 1 total 0.032739 0.028454
2 10000 3 1 total 0.120780 0.170809
3 10000 4 1 total 0.302780 0.109110
4 10000 5 1 total 0.000000 0.000000
5 10000 6 1 total 0.000000 0.000000
0 10000 1 1 total 0.013355 0.001207
1 10000 2 1 total 0.032600 0.002866
2 10000 3 1 total 0.121083 0.010442
3 10000 4 1 total 0.305910 0.025627
4 10000 5 1 total 0.861934 0.068282
5 10000 6 1 total 2.895065 0.230223
material delayedgroup group in group out nuclide mean std. dev.
0 10000 1 1 1 total 0.000000 0.000000
1 10000 2 1 1 total 0.000384 0.000236

4
tests/test_mgxs_library_distribcell/inputs_true.dat
View file

@ -423,7 +423,7 @@
<filter bins="0.0 20000000.0" type="energy" />
<nuclides>total</nuclides>
<scores>delayed-nu-fission</scores>
<estimator>analog</estimator>
<estimator>tracklength</estimator>
</tally>
<tally id="10048">
<filter bins="10000" type="distribcell" />
@ -431,7 +431,7 @@
<filter bins="0.0 20000000.0" type="energy" />
<nuclides>total</nuclides>
<scores>decay-rate</scores>
<estimator>analog</estimator>
<estimator>tracklength</estimator>
</tally>
<tally id="10049">
<filter bins="10000" type="distribcell" />

12
tests/test_mgxs_library_distribcell/results_true.dat
View file

@ -64,12 +64,12 @@
4 ((0, 1, 2, 3, 4, 5, 6, 7, 8, 9, 10, 11, 12, 13... 5 1 total 0.001210 0.000058
5 ((0, 1, 2, 3, 4, 5, 6, 7, 8, 9, 10, 11, 12, 13... 6 1 total 0.000504 0.000024
sum(distribcell) delayedgroup group in nuclide mean std. dev.
0 ((0, 1, 2, 3, 4, 5, 6, 7, 8, 9, 10, 11, 12, 13... 1 1 total 0.000000 0.000000
1 ((0, 1, 2, 3, 4, 5, 6, 7, 8, 9, 10, 11, 12, 13... 2 1 total 0.032739 0.046300
2 ((0, 1, 2, 3, 4, 5, 6, 7, 8, 9, 10, 11, 12, 13... 3 1 total 0.120780 0.170809
3 ((0, 1, 2, 3, 4, 5, 6, 7, 8, 9, 10, 11, 12, 13... 4 1 total 0.000000 0.000000
4 ((0, 1, 2, 3, 4, 5, 6, 7, 8, 9, 10, 11, 12, 13... 5 1 total 0.000000 0.000000
5 ((0, 1, 2, 3, 4, 5, 6, 7, 8, 9, 10, 11, 12, 13... 6 1 total 2.853000 4.034751
0 ((0, 1, 2, 3, 4, 5, 6, 7, 8, 9, 10, 11, 12, 13... 1 1 total 0.013353 0.000686
1 ((0, 1, 2, 3, 4, 5, 6, 7, 8, 9, 10, 11, 12, 13... 2 1 total 0.032613 0.001627
2 ((0, 1, 2, 3, 4, 5, 6, 7, 8, 9, 10, 11, 12, 13... 3 1 total 0.121054 0.005911
3 ((0, 1, 2, 3, 4, 5, 6, 7, 8, 9, 10, 11, 12, 13... 4 1 total 0.305627 0.014428
4 ((0, 1, 2, 3, 4, 5, 6, 7, 8, 9, 10, 11, 12, 13... 5 1 total 0.860892 0.037879
5 ((0, 1, 2, 3, 4, 5, 6, 7, 8, 9, 10, 11, 12, 13... 6 1 total 2.891521 0.127879
sum(distribcell) delayedgroup group in group out nuclide mean std. dev.
0 ((0, 1, 2, 3, 4, 5, 6, 7, 8, 9, 10, 11, 12, 13... 1 1 1 total 0.000000 0.000000
1 ((0, 1, 2, 3, 4, 5, 6, 7, 8, 9, 10, 11, 12, 13... 2 1 1 total 0.000175 0.000175

12
tests/test_mgxs_library_hdf5/inputs_true.dat
View file

@ -396,7 +396,7 @@
<filter bins="0.0 0.625 20000000.0" type="energy" />
<nuclides>total</nuclides>
<scores>delayed-nu-fission</scores>
<estimator>analog</estimator>
<estimator>tracklength</estimator>
</tally>
<tally id="10048">
<filter bins="10000" type="material" />
@ -404,7 +404,7 @@
<filter bins="0.0 0.625 20000000.0" type="energy" />
<nuclides>total</nuclides>
<scores>decay-rate</scores>
<estimator>analog</estimator>
<estimator>tracklength</estimator>
</tally>
<tally id="10049">
<filter bins="10000" type="material" />
@ -767,7 +767,7 @@
<filter bins="0.0 0.625 20000000.0" type="energy" />
<nuclides>total</nuclides>
<scores>delayed-nu-fission</scores>
<estimator>analog</estimator>
<estimator>tracklength</estimator>
</tally>
<tally id="10099">
<filter bins="10001" type="material" />
@ -775,7 +775,7 @@
<filter bins="0.0 0.625 20000000.0" type="energy" />
<nuclides>total</nuclides>
<scores>decay-rate</scores>
<estimator>analog</estimator>
<estimator>tracklength</estimator>
</tally>
<tally id="10100">
<filter bins="10001" type="material" />
@ -1138,7 +1138,7 @@
<filter bins="0.0 0.625 20000000.0" type="energy" />
<nuclides>total</nuclides>
<scores>delayed-nu-fission</scores>
<estimator>analog</estimator>
<estimator>tracklength</estimator>
</tally>
<tally id="10150">
<filter bins="10002" type="material" />
@ -1146,7 +1146,7 @@
<filter bins="0.0 0.625 20000000.0" type="energy" />
<nuclides>total</nuclides>
<scores>decay-rate</scores>
<estimator>analog</estimator>
<estimator>tracklength</estimator>
</tally>
<tally id="10151">
<filter bins="10002" type="material" />

24
tests/test_mgxs_library_hdf5/results_true.dat
View file

@ -117,18 +117,18 @@ domain=10000 type=beta
[1.82980497e-04 4.50738567e-04]
[7.48899920e-05 1.88812772e-04]]
domain=10000 type=decay-rate
[[0.00000000e+00 0.00000000e+00]
[0.00000000e+00 3.27390000e-02]
[0.00000000e+00 1.20780000e-01]
[0.00000000e+00 3.02780000e-01]
[0.00000000e+00 0.00000000e+00]
[0.00000000e+00 0.00000000e+00]]
[[0.00000000e+00 0.00000000e+00]
[0.00000000e+00 2.84543737e-02]
[0.00000000e+00 1.70808714e-01]
[0.00000000e+00 1.09109511e-01]
[0.00000000e+00 0.00000000e+00]
[0.00000000e+00 0.00000000e+00]]
[[1.34450193e-02 1.33360001e-02]
[3.20638662e-02 3.27389978e-02]
[1.22136025e-01 1.20780007e-01]
[3.15269337e-01 3.02780066e-01]
[8.89232590e-01 8.49490287e-01]
[2.98940410e+00 2.85300088e+00]]
[[1.08439397e-03 1.44623816e-03]
[2.65794939e-03 3.55041884e-03]
[1.02955008e-02 1.30981285e-02]
[2.71748525e-02 3.28353351e-02]
[7.93682876e-02 9.21239479e-02]
[2.66253276e-01 3.09396954e-01]]
domain=10000 type=delayed-nu-fission matrix
[[[0.00000000e+00 0.00000000e+00]
[0.00000000e+00 0.00000000e+00]]

4
tests/test_mgxs_library_mesh/inputs_true.dat
View file

@ -660,7 +660,7 @@
<filter bins="0.0 20000000.0" type="energy" />
<nuclides>total</nuclides>
<scores>delayed-nu-fission</scores>
<estimator>analog</estimator>
<estimator>tracklength</estimator>
</tally>
<tally id="10048">
<filter bins="1" type="mesh" />
@ -668,7 +668,7 @@
<filter bins="0.0 20000000.0" type="energy" />
<nuclides>total</nuclides>
<scores>decay-rate</scores>
<estimator>analog</estimator>
<estimator>tracklength</estimator>
</tally>
<tally id="10049">
<filter bins="1" type="mesh" />

52
tests/test_mgxs_library_mesh/results_true.dat
View file

@ -214,32 +214,32 @@
21 2 2 1 4 1 total 0.002675 0.000783
22 2 2 1 5 1 total 0.001199 0.000341
23 2 2 1 6 1 total 0.000499 0.000142
mesh 1 delayedgroup group in nuclide mean std. dev.
x y z
0 1 1 1 1 1 total 0.00000 0.00000
1 1 1 1 2 1 total 0.00000 0.00000
2 1 1 1 3 1 total 0.00000 0.00000
3 1 1 1 4 1 total 0.00000 0.00000
4 1 1 1 5 1 total 0.84949 1.20136
5 1 1 1 6 1 total 0.00000 0.00000
6 1 2 1 1 1 total 0.00000 0.00000
7 1 2 1 2 1 total 0.00000 0.00000
8 1 2 1 3 1 total 0.00000 0.00000
9 1 2 1 4 1 total 0.00000 0.00000
10 1 2 1 5 1 total 0.00000 0.00000
11 1 2 1 6 1 total 0.00000 0.00000
12 2 1 1 1 1 total 0.00000 0.00000
13 2 1 1 2 1 total 0.00000 0.00000
14 2 1 1 3 1 total 0.00000 0.00000
15 2 1 1 4 1 total 0.00000 0.00000
16 2 1 1 5 1 total 0.00000 0.00000
17 2 1 1 6 1 total 0.00000 0.00000
18 2 2 1 1 1 total 0.00000 0.00000
19 2 2 1 2 1 total 0.00000 0.00000
20 2 2 1 3 1 total 0.00000 0.00000
21 2 2 1 4 1 total 0.00000 0.00000
22 2 2 1 5 1 total 0.00000 0.00000
23 2 2 1 6 1 total 0.00000 0.00000
mesh 1 delayedgroup group in nuclide mean std. dev.
x y z
0 1 1 1 1 1 total 0.013362 0.003586
1 1 1 1 2 1 total 0.032554 0.008644
2 1 1 1 3 1 total 0.121179 0.031941
3 1 1 1 4 1 total 0.306858 0.079970
4 1 1 1 5 1 total 0.865303 0.220416
5 1 1 1 6 1 total 2.906554 0.741587
6 1 2 1 1 1 total 0.013362 0.004521
7 1 2 1 2 1 total 0.032556 0.011045
8 1 2 1 3 1 total 0.121174 0.041226
9 1 2 1 4 1 total 0.306816 0.104995
10 1 2 1 5 1 total 0.865155 0.301124
11 1 2 1 6 1 total 2.906049 1.010025
12 2 1 1 1 1 total 0.013353 0.002006
13 2 1 1 2 1 total 0.032614 0.004677
14 2 1 1 3 1 total 0.121052 0.016785
15 2 1 1 4 1 total 0.305600 0.040199
16 2 1 1 5 1 total 0.860792 0.101713
17 2 1 1 6 1 total 2.891182 0.344217
18 2 2 1 1 1 total 0.013356 0.004047
19 2 2 1 2 1 total 0.032596 0.009453
20 2 2 1 3 1 total 0.121091 0.034091
21 2 2 1 4 1 total 0.305990 0.082530
22 2 2 1 5 1 total 0.862223 0.216936
23 2 2 1 6 1 total 2.896051 0.731528
mesh 1 delayedgroup group in group out nuclide mean std. dev.
x y z
0 1 1 1 1 1 1 total 0.000000 0.000000

12
tests/test_mgxs_library_no_nuclides/inputs_true.dat
View file

@ -396,7 +396,7 @@
<filter bins="0.0 0.625 20000000.0" type="energy" />
<nuclides>total</nuclides>
<scores>delayed-nu-fission</scores>
<estimator>analog</estimator>
<estimator>tracklength</estimator>
</tally>
<tally id="10048">
<filter bins="10000" type="material" />
@ -404,7 +404,7 @@
<filter bins="0.0 0.625 20000000.0" type="energy" />
<nuclides>total</nuclides>
<scores>decay-rate</scores>
<estimator>analog</estimator>
<estimator>tracklength</estimator>
</tally>
<tally id="10049">
<filter bins="10000" type="material" />
@ -767,7 +767,7 @@
<filter bins="0.0 0.625 20000000.0" type="energy" />
<nuclides>total</nuclides>
<scores>delayed-nu-fission</scores>
<estimator>analog</estimator>
<estimator>tracklength</estimator>
</tally>
<tally id="10099">
<filter bins="10001" type="material" />
@ -775,7 +775,7 @@
<filter bins="0.0 0.625 20000000.0" type="energy" />
<nuclides>total</nuclides>
<scores>decay-rate</scores>
<estimator>analog</estimator>
<estimator>tracklength</estimator>
</tally>
<tally id="10100">
<filter bins="10001" type="material" />
@ -1138,7 +1138,7 @@
<filter bins="0.0 0.625 20000000.0" type="energy" />
<nuclides>total</nuclides>
<scores>delayed-nu-fission</scores>
<estimator>analog</estimator>
<estimator>tracklength</estimator>
</tally>
<tally id="10150">
<filter bins="10002" type="material" />
@ -1146,7 +1146,7 @@
<filter bins="0.0 0.625 20000000.0" type="energy" />
<nuclides>total</nuclides>
<scores>decay-rate</scores>
<estimator>analog</estimator>
<estimator>tracklength</estimator>
</tally>
<tally id="10151">
<filter bins="10002" type="material" />

24
tests/test_mgxs_library_no_nuclides/results_true.dat
View file

@ -129,18 +129,18 @@
8 10000 5 2 total 0.004684 0.000451
10 10000 6 2 total 0.001962 0.000189
material delayedgroup group in nuclide mean std. dev.
1 10000 1 1 total 0.000000 0.000000
3 10000 2 1 total 0.000000 0.000000
5 10000 3 1 total 0.000000 0.000000
7 10000 4 1 total 0.000000 0.000000
9 10000 5 1 total 0.000000 0.000000
11 10000 6 1 total 0.000000 0.000000
0 10000 1 2 total 0.000000 0.000000
2 10000 2 2 total 0.032739 0.028454
4 10000 3 2 total 0.120780 0.170809
6 10000 4 2 total 0.302780 0.109110
8 10000 5 2 total 0.000000 0.000000
10 10000 6 2 total 0.000000 0.000000
1 10000 1 1 total 0.013445 0.001084
3 10000 2 1 total 0.032064 0.002658
5 10000 3 1 total 0.122136 0.010296
7 10000 4 1 total 0.315269 0.027175
9 10000 5 1 total 0.889233 0.079368
11 10000 6 1 total 2.989404 0.266253
0 10000 1 2 total 0.013336 0.001446
2 10000 2 2 total 0.032739 0.003550
4 10000 3 2 total 0.120780 0.013098
6 10000 4 2 total 0.302780 0.032835
8 10000 5 2 total 0.849490 0.092124
10 10000 6 2 total 2.853001 0.309397
material delayedgroup group in group out nuclide mean std. dev.
3 10000 1 1 1 total 0.000000 0.000000
7 10000 2 1 1 total 0.000000 0.000000