crp/OpenMC
1
0
Fork 0
mirror of https://github.com/openmc-dev/openmc.git synced 2026-07-28 06:05:58 -04:00
Code Issues Projects Releases Packages Wiki Activity Actions 5

Merge remote-tracking branch 'upstream/develop' into hotfix-scatt-mat-nuclides

Browse source
This commit is contained in:
wboyd 2017-04-11 10:25:27 -04:00
commit 7673ea222e
121 changed files with 1816 additions and 1831 deletions
Download patch file Download diff file Expand all files Collapse all files

24
CMakeLists.txt
View file

@ -25,14 +25,13 @@ endif()
# Command line options
#===============================================================================
option(openmp "Enable shared-memory parallelism with OpenMP" OFF)
option(openmp "Enable shared-memory parallelism with OpenMP" ON)
option(profile "Compile with profiling flags" OFF)
option(debug "Compile with debug flags" OFF)
option(optimize "Turn on all compiler optimization flags" OFF)
option(coverage "Compile with coverage analysis flags" OFF)
option(mpif08 "Use Fortran 2008 MPI interface" OFF)
# Maximum number of nested coordinates levels
set(maxcoord 10 CACHE STRING "Maximum number of nested coordinate levels")
add_definitions(-DMAX_COORD=${maxcoord})
@ -108,9 +107,11 @@ if(CMAKE_Fortran_COMPILER_ID STREQUAL GNU)
endif()
# GCC compiler options
list(APPEND f90flags -cpp -std=f2008 -fbacktrace)
list(APPEND cflags -cpp -std=c99)
list(APPEND f90flags -cpp -std=f2008 -fbacktrace -O2)
list(APPEND cflags -cpp -std=c99 -O2)
if(debug)
list(REMOVE_ITEM f90flags -O2)
list(REMOVE_ITEM cflags -O2)
list(APPEND f90flags -g -Wall -pedantic -fbounds-check
-ffpe-trap=invalid,overflow,underflow)
list(APPEND cflags -g -Wall -pedantic -fbounds-check)
@ -122,6 +123,8 @@ if(CMAKE_Fortran_COMPILER_ID STREQUAL GNU)
list(APPEND ldflags -pg)
endif()
if(optimize)
list(REMOVE_ITEM f90flags -O2)
list(REMOVE_ITEM cflags -O2)
list(APPEND f90flags -O3)
list(APPEND cflags -O3)
endif()
@ -142,8 +145,8 @@ elseif(CMAKE_Fortran_COMPILER_ID STREQUAL Intel)
list(APPEND cflags -std=c99)
if(debug)
list(APPEND f90flags -g -warn -ftrapuv -fp-stack-check
"-check all" -fpe0)
list(APPEND cflags -g -w3 -ftrapuv -fp-stack-check)
"-check all" -fpe0 -O0)
list(APPEND cflags -g -w3 -ftrapuv -fp-stack-check -O0)
list(APPEND ldflags -g)
endif()
if(profile)
@ -182,7 +185,8 @@ elseif(CMAKE_Fortran_COMPILER_ID STREQUAL XL)
list(APPEND f90flags -O2)
add_definitions(-DNO_F2008)
if(debug)
list(APPEND f90flags -g -C -qflag=i:i -u)
list(REMOVE_ITEM f90flags -O2)
list(APPEND f90flags -g -C -qflag=i:i -u -O0)
list(APPEND ldflags -g)
endif()
if(profile)
@ -190,6 +194,7 @@ elseif(CMAKE_Fortran_COMPILER_ID STREQUAL XL)
list(APPEND ldflags -p)
endif()
if(optimize)
list(REMOVE_ITEM f90flags -O2)
list(APPEND f90flags -O3)
endif()
if(openmp)
@ -207,6 +212,11 @@ elseif(CMAKE_Fortran_COMPILER_ID STREQUAL Cray)
endif()
# Show flags being used
message(STATUS "Fortran flags: ${f90flags}")
message(STATUS "C flags: ${cflags}")
message(STATUS "Linker flags: ${ldflags}")
#===============================================================================
# git SHA1 hash
#===============================================================================

17
docs/source/devguide/workflow.rst
View file

@ -105,15 +105,20 @@ in the tests directory. We recommend to developers to test their branches
before submitting a formal pull request using gfortran and Intel compilers
if available.
The test suite is designed to integrate with cmake using ctest_.
It is configured to run with cross sections from NNDC_. To
download these cross sections please do the following:
The test suite is designed to integrate with cmake using ctest_. It is
configured to run with cross sections from NNDC_ augmented with 0 K elastic
scattering data for select nuclides as well as multipole data. To download the
proper data, run the following commands:
.. code-block:: sh
cd ../scripts
./openmc-get-nndc-data
export OPENMC_CROSS_SECTIONS=<path_to_data_folder>/nndc_hdf5/cross_sections.xml
wget -O nndc_hdf5.tar.xz $(cat <openmc_root>/.travis.yml | grep anl.box | awk '{print $2}')
tar xJvf nndc_hdf5.tar.xz
export OPENMC_CROSS_SECTIONS=$(pwd)/nndc_hdf5/cross_sections.xml
git clone --branch=master git://github.com/smharper/windowed_multipole_library.git wmp_lib
tar xzvf wmp_lib/multipole_lib.tar.gz
export OPENMC_MULTIPOLE_LIBRARY=$(pwd)/multipole_lib
The test suite can be run on an already existing build using:

25
docs/source/pythonapi/examples.rst Normal file
View file

@ -0,0 +1,25 @@
----------------------------------------
:mod:`openmc.examples` -- Example Models
----------------------------------------
Simple Models
-------------
.. autosummary::
:toctree: generated
:nosignatures:
:template: myfunction.rst
openmc.examples.slab_mg
Reactor Models
--------------
.. autosummary::
:toctree: generated
:nosignatures:
:template: myfunction.rst
openmc.examples.pwr_pin_cell
openmc.examples.pwr_assembly
openmc.examples.pwr_core

1
docs/source/pythonapi/index.rst
View file

@ -47,4 +47,5 @@ Modules
mgxs
model
data
examples
openmoc

24
docs/source/usersguide/cross_sections.rst
View file

@ -199,6 +199,30 @@ OpenMC.
etc. For a more thorough overview of the capabilities of this class,
see the :ref:`notebook_nuclear_data` example notebook.
Enabling Resonance Scattering Treatments
----------------------------------------
In order for OpenMC to correctly treat elastic scattering in heavy nuclides
where low-lying resonances might be present (see
:ref:`energy_dependent_xs_model`), the elastic scattering cross section at 0 K
must be present. To add the 0 K elastic scattering cross section to existing
:class:`IncidentNeutron` instance, you can use the
:meth:`IncidentNeutron.add_elastic_0K_from_endf` method which requires an ENDF
file for the nuclide you are modifying::
u238 = openmc.data.IncidentNeutron.from_hdf5('U238.h5')
u238.add_elastic_0K_from_endf('n-092_U_238.endf')
u238.export_to_hdf5('U238_with_0K.h5')
With 0 K elastic scattering data present, you can turn on a resonance scattering
method using :attr:`Settings.resonance_scattering`.
.. note:: The process of reconstructing resonances and generating tabulated 0 K
cross sections can be computationally expensive, especially for
nuclides like U-238 where thousands of resonances are present. Thus,
running the :meth:`IncidentNeutron.add_elastic_0K_from_endf` method
may take several minutes to complete.
-----------------------
Windowed Multipole Data
-----------------------

2
docs/source/usersguide/install.rst
View file

@ -231,7 +231,7 @@ optimize
openmp
Enables shared-memory parallelism using the OpenMP API. The Fortran compiler
being used must support OpenMP.
being used must support OpenMP. (Default: on)
coverage
Compile and link code instrumented for coverage analysis. This is typically

39
openmc/cell.py
View file

@ -1,4 +1,5 @@
from collections import OrderedDict, Iterable
from copy import deepcopy
from math import cos, sin, pi
from numbers import Real, Integral
from xml.etree import ElementTree as ET
@ -504,6 +505,44 @@ class Cell(object):
return universes
def clone(self, memo=None):
"""Create a copy of this cell with a new unique ID, and clones
the cell's region and fill.
Parameters
----------
memo : dict or None
A nested dictionary of previously cloned objects. This parameter
is used internally and should not be specified by the user.
Returns
-------
clone : openmc.Cell
The clone of this cell
"""
if memo is None:
memo = {}
# If no nemoize'd clone exists, instantiate one
if self not in memo:
clone = deepcopy(self)
clone.id = None
if self.region is not None:
clone.region = self.region.clone(memo)
if self.fill is not None:
if self.fill_type == 'distribmat':
clone.fill = [fill.clone(memo) if fill is not None else None
for fill in self.fill]
else:
clone.fill = self.fill.clone(memo)
# Memoize the clone
memo[self] = clone
return memo[self]
def create_xml_subelement(self, xml_element):
element = ET.Element("cell")
element.set("id", str(self.id))

629
openmc/examples.py Normal file
View file

@ -0,0 +1,629 @@
import numpy as np
import openmc
import openmc.model
def pwr_pin_cell():
"""Create a PWR pin-cell model.
This model is a single fuel pin with 2.4 w/o enriched UO2 corresponding to a
beginning-of-cycle condition and borated water. The specifications are from
the `BEAVRS <http://crpg.mit.edu/research/beavrs>`_ benchmark. Note that the
number of particles/batches is initially set very low for testing purposes.
Returns
-------
model : openmc.model.Model
A PWR pin-cell model
"""
model = openmc.model.Model()
# Define materials.
fuel = openmc.Material(name='UO2 (2.4%)')
fuel.set_density('g/cm3', 10.29769)
fuel.add_nuclide("U234", 4.4843e-6)
fuel.add_nuclide("U235", 5.5815e-4)
fuel.add_nuclide("U238", 2.2408e-2)
fuel.add_nuclide("O16", 4.5829e-2)
clad = openmc.Material(name='Zircaloy')
clad.set_density('g/cm3', 6.55)
clad.add_nuclide("Zr90", 2.1827e-2)
clad.add_nuclide("Zr91", 4.7600e-3)
clad.add_nuclide("Zr92", 7.2758e-3)
clad.add_nuclide("Zr94", 7.3734e-3)
clad.add_nuclide("Zr96", 1.1879e-3)
hot_water = openmc.Material(name='Hot borated water')
hot_water.set_density('g/cm3', 0.740582)
hot_water.add_nuclide("H1", 4.9457e-2)
hot_water.add_nuclide("O16", 2.4672e-2)
hot_water.add_nuclide("B10", 8.0042e-6)
hot_water.add_nuclide("B11", 3.2218e-5)
hot_water.add_s_alpha_beta('c_H_in_H2O')
# Define the materials file.
model.materials = (fuel, clad, hot_water)
# Instantiate ZCylinder surfaces
pitch = 1.26
fuel_or = openmc.ZCylinder(x0=0, y0=0, R=0.39218, name='Fuel OR')
clad_or = openmc.ZCylinder(x0=0, y0=0, R=0.45720, name='Clad OR')
left = openmc.XPlane(x0=-pitch/2, name='left', boundary_type='reflective')
right = openmc.XPlane(x0=pitch/2, name='right', boundary_type='reflective')
bottom = openmc.YPlane(y0=-pitch/2, name='bottom',
boundary_type='reflective')
top = openmc.YPlane(y0=pitch/2, name='top', boundary_type='reflective')
# Instantiate Cells
fuel_pin = openmc.Cell(name='Fuel', fill=fuel)
cladding = openmc.Cell(name='Cladding', fill=clad)
water = openmc.Cell(name='Water', fill=hot_water)
# Use surface half-spaces to define regions
fuel_pin.region = -fuel_or
cladding.region = +fuel_or & -clad_or
water.region = +clad_or & +left & -right & +bottom & -top
# Create root universe
model.geometry.root_universe = openmc.Universe(0, name='root universe')
model.geometry.root_universe.add_cells([fuel_pin, cladding, water])
model.settings.batches = 10
model.settings.inactive = 5
model.settings.particles = 100
model.settings.source = openmc.Source(space=openmc.stats.Box(
[-pitch/2, -pitch/2, -1], [pitch/2, pitch/2, 1], only_fissionable=True))
plot = openmc.Plot.from_geometry(model.geometry)
plot.pixels = (300, 300)
plot.color_by = 'material'
model.plots.append(plot)
return model
def pwr_core():
"""Create a PWR full-core model.
This model is the OECD/NEA Monte Carlo Performance benchmark which is a
grossly simplified pressurized water reactor (PWR) with 241 fuel
assemblies. Note that the number of particles/batches is initially set very
low for testing purposes.
Returns
-------
model : openmc.model.Model
Full-core PWR model
"""
model = openmc.model.Model()
# Define materials.
fuel = openmc.Material(1, name='UOX fuel')
fuel.set_density('g/cm3', 10.062)
fuel.add_nuclide("U234", 4.9476e-6)
fuel.add_nuclide("U235", 4.8218e-4)
fuel.add_nuclide("U238", 2.1504e-2)
fuel.add_nuclide("Xe135", 1.0801e-8)
fuel.add_nuclide("O16", 4.5737e-2)
clad = openmc.Material(2, name='Zircaloy')
clad.set_density('g/cm3', 5.77)
clad.add_nuclide("Zr90", 0.5145)
clad.add_nuclide("Zr91", 0.1122)
clad.add_nuclide("Zr92", 0.1715)
clad.add_nuclide("Zr94", 0.1738)
clad.add_nuclide("Zr96", 0.0280)
cold_water = openmc.Material(3, name='Cold borated water')
cold_water.set_density('atom/b-cm', 0.07416)
cold_water.add_nuclide("H1", 2.0)
cold_water.add_nuclide("O16", 1.0)
cold_water.add_nuclide("B10", 6.490e-4)
cold_water.add_nuclide("B11", 2.689e-3)
cold_water.add_s_alpha_beta('c_H_in_H2O')
hot_water = openmc.Material(4, name='Hot borated water')
hot_water.set_density('atom/b-cm', 0.06614)
hot_water.add_nuclide("H1", 2.0)
hot_water.add_nuclide("O16", 1.0)
hot_water.add_nuclide("B10", 6.490e-4)
hot_water.add_nuclide("B11", 2.689e-3)
hot_water.add_s_alpha_beta('c_H_in_H2O')
rpv_steel = openmc.Material(5, name='Reactor pressure vessel steel')
rpv_steel.set_density('g/cm3', 7.9)
rpv_steel.add_nuclide("Fe54", 0.05437098, 'wo')
rpv_steel.add_nuclide("Fe56", 0.88500663, 'wo')
rpv_steel.add_nuclide("Fe57", 0.0208008, 'wo')
rpv_steel.add_nuclide("Fe58", 0.00282159, 'wo')
rpv_steel.add_nuclide("Ni58", 0.0067198, 'wo')
rpv_steel.add_nuclide("Ni60", 0.0026776, 'wo')
rpv_steel.add_nuclide("Mn55", 0.01, 'wo')
rpv_steel.add_nuclide("Cr52", 0.002092475, 'wo')
rpv_steel.add_nuclide("C0", 0.0025, 'wo')
rpv_steel.add_nuclide("Cu63", 0.0013696, 'wo')
lower_rad_ref = openmc.Material(6, name='Lower radial reflector')
lower_rad_ref.set_density('g/cm3', 4.32)
lower_rad_ref.add_nuclide("H1", 0.0095661, 'wo')
lower_rad_ref.add_nuclide("O16", 0.0759107, 'wo')
lower_rad_ref.add_nuclide("B10", 3.08409e-5, 'wo')
lower_rad_ref.add_nuclide("B11", 1.40499e-4, 'wo')
lower_rad_ref.add_nuclide("Fe54", 0.035620772088, 'wo')
lower_rad_ref.add_nuclide("Fe56", 0.579805982228, 'wo')
lower_rad_ref.add_nuclide("Fe57", 0.01362750048, 'wo')
lower_rad_ref.add_nuclide("Fe58", 0.001848545204, 'wo')
lower_rad_ref.add_nuclide("Ni58", 0.055298376566, 'wo')
lower_rad_ref.add_nuclide("Mn55", 0.0182870, 'wo')
lower_rad_ref.add_nuclide("Cr52", 0.145407678031, 'wo')
lower_rad_ref.add_s_alpha_beta('c_H_in_H2O')
upper_rad_ref = openmc.Material(7, name='Upper radial reflector / Top plate region')
upper_rad_ref.set_density('g/cm3', 4.28)
upper_rad_ref.add_nuclide("H1", 0.0086117, 'wo')
upper_rad_ref.add_nuclide("O16", 0.0683369, 'wo')
upper_rad_ref.add_nuclide("B10", 2.77638e-5, 'wo')
upper_rad_ref.add_nuclide("B11", 1.26481e-4, 'wo')
upper_rad_ref.add_nuclide("Fe54", 0.035953677186, 'wo')
upper_rad_ref.add_nuclide("Fe56", 0.585224740891, 'wo')
upper_rad_ref.add_nuclide("Fe57", 0.01375486056, 'wo')
upper_rad_ref.add_nuclide("Fe58", 0.001865821363, 'wo')
upper_rad_ref.add_nuclide("Ni58", 0.055815129186, 'wo')
upper_rad_ref.add_nuclide("Mn55", 0.0184579, 'wo')
upper_rad_ref.add_nuclide("Cr52", 0.146766614995, 'wo')
upper_rad_ref.add_s_alpha_beta('c_H_in_H2O')
bot_plate = openmc.Material(8, name='Bottom plate region')
bot_plate.set_density('g/cm3', 7.184)
bot_plate.add_nuclide("H1", 0.0011505, 'wo')
bot_plate.add_nuclide("O16", 0.0091296, 'wo')
bot_plate.add_nuclide("B10", 3.70915e-6, 'wo')
bot_plate.add_nuclide("B11", 1.68974e-5, 'wo')
bot_plate.add_nuclide("Fe54", 0.03855611055, 'wo')
bot_plate.add_nuclide("Fe56", 0.627585036425, 'wo')
bot_plate.add_nuclide("Fe57", 0.014750478, 'wo')
bot_plate.add_nuclide("Fe58", 0.002000875025, 'wo')
bot_plate.add_nuclide("Ni58", 0.059855207342, 'wo')
bot_plate.add_nuclide("Mn55", 0.0197940, 'wo')
bot_plate.add_nuclide("Cr52", 0.157390026871, 'wo')
bot_plate.add_s_alpha_beta('c_H_in_H2O')
bot_nozzle = openmc.Material(9, name='Bottom nozzle region')
bot_nozzle.set_density('g/cm3', 2.53)
bot_nozzle.add_nuclide("H1", 0.0245014, 'wo')
bot_nozzle.add_nuclide("O16", 0.1944274, 'wo')
bot_nozzle.add_nuclide("B10", 7.89917e-5, 'wo')
bot_nozzle.add_nuclide("B11", 3.59854e-4, 'wo')
bot_nozzle.add_nuclide("Fe54", 0.030411411144, 'wo')
bot_nozzle.add_nuclide("Fe56", 0.495012237964, 'wo')
bot_nozzle.add_nuclide("Fe57", 0.01163454624, 'wo')
bot_nozzle.add_nuclide("Fe58", 0.001578204652, 'wo')
bot_nozzle.add_nuclide("Ni58", 0.047211231662, 'wo')
bot_nozzle.add_nuclide("Mn55", 0.0156126, 'wo')
bot_nozzle.add_nuclide("Cr52", 0.124142524198, 'wo')
bot_nozzle.add_s_alpha_beta('c_H_in_H2O')
top_nozzle = openmc.Material(10, name='Top nozzle region')
top_nozzle.set_density('g/cm3', 1.746)
top_nozzle.add_nuclide("H1", 0.0358870, 'wo')
top_nozzle.add_nuclide("O16", 0.2847761, 'wo')
top_nozzle.add_nuclide("B10", 1.15699e-4, 'wo')
top_nozzle.add_nuclide("B11", 5.27075e-4, 'wo')
top_nozzle.add_nuclide("Fe54", 0.02644016154, 'wo')
top_nozzle.add_nuclide("Fe56", 0.43037146399, 'wo')
top_nozzle.add_nuclide("Fe57", 0.0101152584, 'wo')
top_nozzle.add_nuclide("Fe58", 0.00137211607, 'wo')
top_nozzle.add_nuclide("Ni58", 0.04104621835, 'wo')
top_nozzle.add_nuclide("Mn55", 0.0135739, 'wo')
top_nozzle.add_nuclide("Cr52", 0.107931450781, 'wo')
top_nozzle.add_s_alpha_beta('c_H_in_H2O')
top_fa = openmc.Material(11, name='Top of fuel assemblies')
top_fa.set_density('g/cm3', 3.044)
top_fa.add_nuclide("H1", 0.0162913, 'wo')
top_fa.add_nuclide("O16", 0.1292776, 'wo')
top_fa.add_nuclide("B10", 5.25228e-5, 'wo')
top_fa.add_nuclide("B11", 2.39272e-4, 'wo')
top_fa.add_nuclide("Zr90", 0.43313403903, 'wo')
top_fa.add_nuclide("Zr91", 0.09549277374, 'wo')
top_fa.add_nuclide("Zr92", 0.14759527104, 'wo')
top_fa.add_nuclide("Zr94", 0.15280552077, 'wo')
top_fa.add_nuclide("Zr96", 0.02511169542, 'wo')
top_fa.add_s_alpha_beta('c_H_in_H2O')
bot_fa = openmc.Material(12, name='Bottom of fuel assemblies')
bot_fa.set_density('g/cm3', 1.762)
bot_fa.add_nuclide("H1", 0.0292856, 'wo')
bot_fa.add_nuclide("O16", 0.2323919, 'wo')
bot_fa.add_nuclide("B10", 9.44159e-5, 'wo')
bot_fa.add_nuclide("B11", 4.30120e-4, 'wo')
bot_fa.add_nuclide("Zr90", 0.3741373658, 'wo')
bot_fa.add_nuclide("Zr91", 0.0824858164, 'wo')
bot_fa.add_nuclide("Zr92", 0.1274914944, 'wo')
bot_fa.add_nuclide("Zr94", 0.1319920622, 'wo')
bot_fa.add_nuclide("Zr96", 0.0216912612, 'wo')
bot_fa.add_s_alpha_beta('c_H_in_H2O')
# Define the materials file.
model.materials = (fuel, clad, cold_water, hot_water, rpv_steel,
lower_rad_ref, upper_rad_ref, bot_plate,
bot_nozzle, top_nozzle, top_fa, bot_fa)
# Define surfaces.
s1 = openmc.ZCylinder(R=0.41, surface_id=1)
s2 = openmc.ZCylinder(R=0.475, surface_id=2)
s3 = openmc.ZCylinder(R=0.56, surface_id=3)
s4 = openmc.ZCylinder(R=0.62, surface_id=4)
s5 = openmc.ZCylinder(R=187.6, surface_id=5)
s6 = openmc.ZCylinder(R=209.0, surface_id=6)
s7 = openmc.ZCylinder(R=229.0, surface_id=7)
s8 = openmc.ZCylinder(R=249.0, surface_id=8, boundary_type='vacuum')
s31 = openmc.ZPlane(z0=-229.0, surface_id=31, boundary_type='vacuum')
s32 = openmc.ZPlane(z0=-199.0, surface_id=32)
s33 = openmc.ZPlane(z0=-193.0, surface_id=33)
s34 = openmc.ZPlane(z0=-183.0, surface_id=34)
s35 = openmc.ZPlane(z0=0.0, surface_id=35)
s36 = openmc.ZPlane(z0=183.0, surface_id=36)
s37 = openmc.ZPlane(z0=203.0, surface_id=37)
s38 = openmc.ZPlane(z0=215.0, surface_id=38)
s39 = openmc.ZPlane(z0=223.0, surface_id=39, boundary_type='vacuum')
# Define pin cells.
fuel_cold = openmc.Universe(name='Fuel pin, cladding, cold water',
universe_id=1)
c21 = openmc.Cell(cell_id=21, fill=fuel, region=-s1)
c22 = openmc.Cell(cell_id=22, fill=clad, region=+s1 & -s2)
c23 = openmc.Cell(cell_id=23, fill=cold_water, region=+s2)
fuel_cold.add_cells((c21, c22, c23))
tube_cold = openmc.Universe(name='Instrumentation guide tube, '
'cold water', universe_id=2)
c24 = openmc.Cell(cell_id=24, fill=cold_water, region=-s3)
c25 = openmc.Cell(cell_id=25, fill=clad, region=+s3 & -s4)
c26 = openmc.Cell(cell_id=26, fill=cold_water, region=+s4)
tube_cold.add_cells((c24, c25, c26))
fuel_hot = openmc.Universe(name='Fuel pin, cladding, hot water',
universe_id=3)
c27 = openmc.Cell(cell_id=27, fill=fuel, region=-s1)
c28 = openmc.Cell(cell_id=28, fill=clad, region=+s1 & -s2)
c29 = openmc.Cell(cell_id=29, fill=hot_water, region=+s2)
fuel_hot.add_cells((c27, c28, c29))
tube_hot = openmc.Universe(name='Instrumentation guide tube, hot water',
universe_id=4)
c30 = openmc.Cell(cell_id=30, fill=hot_water, region=-s3)
c31 = openmc.Cell(cell_id=31, fill=clad, region=+s3 & -s4)
c32 = openmc.Cell(cell_id=32, fill=hot_water, region=+s4)
tube_hot.add_cells((c30, c31, c32))
# Set positions occupied by guide tubes
tube_x = np.array([5, 8, 11, 3, 13, 2, 5, 8, 11, 14, 2, 5, 8, 11, 14,
2, 5, 8, 11, 14, 3, 13, 5, 8, 11])
tube_y = np.array([2, 2, 2, 3, 3, 5, 5, 5, 5, 5, 8, 8, 8, 8, 8,
11, 11, 11, 11, 11, 13, 13, 14, 14, 14])
# Define fuel lattices.
l100 = openmc.RectLattice(name='Fuel assembly (lower half)', lattice_id=100)
l100.lower_left = (-10.71, -10.71)
l100.pitch = (1.26, 1.26)
l100.universes = np.tile(fuel_cold, (17, 17))
l100.universes[tube_x, tube_y] = tube_cold
l101 = openmc.RectLattice(name='Fuel assembly (upper half)', lattice_id=101)
l101.lower_left = (-10.71, -10.71)
l101.pitch = (1.26, 1.26)
l101.universes = np.tile(fuel_hot, (17, 17))
l101.universes[tube_x, tube_y] = tube_hot
# Define assemblies.
fa_cw = openmc.Universe(name='Water assembly (cold)', universe_id=5)
c50 = openmc.Cell(cell_id=50, fill=cold_water, region=+s34 & -s35)
fa_cw.add_cell(c50)
fa_hw = openmc.Universe(name='Water assembly (hot)', universe_id=7)
c70 = openmc.Cell(cell_id=70, fill=hot_water, region=+s35 & -s36)
fa_hw.add_cell(c70)
fa_cold = openmc.Universe(name='Fuel assembly (cold)', universe_id=6)
c60 = openmc.Cell(cell_id=60, fill=l100, region=+s34 & -s35)
fa_cold.add_cell(c60)
fa_hot = openmc.Universe(name='Fuel assembly (hot)', universe_id=8)
c80 = openmc.Cell(cell_id=80, fill=l101, region=+s35 & -s36)
fa_hot.add_cell(c80)
# Define core lattices
l200 = openmc.RectLattice(name='Core lattice (lower half)', lattice_id=200)
l200.lower_left = (-224.91, -224.91)
l200.pitch = (21.42, 21.42)
l200.universes = [
[fa_cw]*21,
[fa_cw]*21,
[fa_cw]*7 + [fa_cold]*7 + [fa_cw]*7,
[fa_cw]*5 + [fa_cold]*11 + [fa_cw]*5,
[fa_cw]*4 + [fa_cold]*13 + [fa_cw]*4,
[fa_cw]*3 + [fa_cold]*15 + [fa_cw]*3,
[fa_cw]*3 + [fa_cold]*15 + [fa_cw]*3,
[fa_cw]*2 + [fa_cold]*17 + [fa_cw]*2,
[fa_cw]*2 + [fa_cold]*17 + [fa_cw]*2,
[fa_cw]*2 + [fa_cold]*17 + [fa_cw]*2,
[fa_cw]*2 + [fa_cold]*17 + [fa_cw]*2,
[fa_cw]*2 + [fa_cold]*17 + [fa_cw]*2,
[fa_cw]*2 + [fa_cold]*17 + [fa_cw]*2,
[fa_cw]*2 + [fa_cold]*17 + [fa_cw]*2,
[fa_cw]*3 + [fa_cold]*15 + [fa_cw]*3,
[fa_cw]*3 + [fa_cold]*15 + [fa_cw]*3,
[fa_cw]*4 + [fa_cold]*13 + [fa_cw]*4,
[fa_cw]*5 + [fa_cold]*11 + [fa_cw]*5,
[fa_cw]*7 + [fa_cold]*7 + [fa_cw]*7,
[fa_cw]*21,
[fa_cw]*21]
l201 = openmc.RectLattice(name='Core lattice (lower half)', lattice_id=201)
l201.lower_left = (-224.91, -224.91)
l201.pitch = (21.42, 21.42)
l201.universes = [
[fa_hw]*21,
[fa_hw]*21,
[fa_hw]*7 + [fa_hot]*7 + [fa_hw]*7,
[fa_hw]*5 + [fa_hot]*11 + [fa_hw]*5,
[fa_hw]*4 + [fa_hot]*13 + [fa_hw]*4,
[fa_hw]*3 + [fa_hot]*15 + [fa_hw]*3,
[fa_hw]*3 + [fa_hot]*15 + [fa_hw]*3,
[fa_hw]*2 + [fa_hot]*17 + [fa_hw]*2,
[fa_hw]*2 + [fa_hot]*17 + [fa_hw]*2,
[fa_hw]*2 + [fa_hot]*17 + [fa_hw]*2,
[fa_hw]*2 + [fa_hot]*17 + [fa_hw]*2,
[fa_hw]*2 + [fa_hot]*17 + [fa_hw]*2,
[fa_hw]*2 + [fa_hot]*17 + [fa_hw]*2,
[fa_hw]*2 + [fa_hot]*17 + [fa_hw]*2,
[fa_hw]*3 + [fa_hot]*15 + [fa_hw]*3,
[fa_hw]*3 + [fa_hot]*15 + [fa_hw]*3,
[fa_hw]*4 + [fa_hot]*13 + [fa_hw]*4,
[fa_hw]*5 + [fa_hot]*11 + [fa_hw]*5,
[fa_hw]*7 + [fa_hot]*7 + [fa_hw]*7,
[fa_hw]*21,
[fa_hw]*21]
# Define root universe.
root = openmc.Universe(universe_id=0, name='root universe')
c1 = openmc.Cell(cell_id=1, fill=l200, region=-s6 & +s34 & -s35)
c2 = openmc.Cell(cell_id=2, fill=l201, region=-s6 & +s35 & -s36)
c3 = openmc.Cell(cell_id=3, fill=bot_plate, region=-s7 & +s31 & -s32)
c4 = openmc.Cell(cell_id=4, fill=bot_nozzle, region=-s5 & +s32 & -s33)
c5 = openmc.Cell(cell_id=5, fill=bot_fa, region=-s5 & +s33 & -s34)
c6 = openmc.Cell(cell_id=6, fill=top_fa, region=-s5 & +s36 & -s37)
c7 = openmc.Cell(cell_id=7, fill=top_nozzle, region=-s5 & +s37 & -s38)
c8 = openmc.Cell(cell_id=8, fill=upper_rad_ref, region=-s7 & +s38 & -s39)
c9 = openmc.Cell(cell_id=9, fill=bot_nozzle, region=+s6 & -s7 & +s32 & -s38)
c10 = openmc.Cell(cell_id=10, fill=rpv_steel, region=+s7 & -s8 & +s31 & -s39)
c11 = openmc.Cell(cell_id=11, fill=lower_rad_ref, region=+s5 & -s6 & +s32 & -s34)
c12 = openmc.Cell(cell_id=12, fill=upper_rad_ref, region=+s5 & -s6 & +s36 & -s38)
root.add_cells((c1, c2, c3, c4, c5, c6, c7, c8, c9, c10, c11, c12))
# Assign root universe to geometry
model.geometry.root_universe = root
model.settings.batches = 10
model.settings.inactive = 5
model.settings.particles = 100
model.settings.source = openmc.Source(space=openmc.stats.Box(
[-160, -160, -183], [160, 160, 183]))
plot = openmc.Plot()
plot.origin = (125, 125, 0)
plot.width = (250, 250)
plot.pixels = (3000, 3000)
plot.color_by = 'material'
model.plots.append(plot)
return model
def pwr_assembly():
"""Create a PWR assembly model.
This model is a reflected 17x17 fuel assembly from the the `BEAVRS
<http://crpg.mit.edu/research/beavrs>`_ benchmark. The fuel is 2.4 w/o
enriched UO2 corresponding to a beginning-of-cycle condition. Note that the
number of particles/batches is initially set very low for testing purposes.
Returns
-------
model : openmc.model.Model
A PWR assembly model
"""
model = openmc.model.Model()
# Define materials.
fuel = openmc.Material(name='Fuel')
fuel.set_density('g/cm3', 10.29769)
fuel.add_nuclide("U234", 4.4843e-6)
fuel.add_nuclide("U235", 5.5815e-4)
fuel.add_nuclide("U238", 2.2408e-2)
fuel.add_nuclide("O16", 4.5829e-2)
clad = openmc.Material(name='Cladding')
clad.set_density('g/cm3', 6.55)
clad.add_nuclide("Zr90", 2.1827e-2)
clad.add_nuclide("Zr91", 4.7600e-3)
clad.add_nuclide("Zr92", 7.2758e-3)
clad.add_nuclide("Zr94", 7.3734e-3)
clad.add_nuclide("Zr96", 1.1879e-3)
hot_water = openmc.Material(name='Hot borated water')
hot_water.set_density('g/cm3', 0.740582)
hot_water.add_nuclide("H1", 4.9457e-2)
hot_water.add_nuclide("O16", 2.4672e-2)
hot_water.add_nuclide("B10", 8.0042e-6)
hot_water.add_nuclide("B11", 3.2218e-5)
hot_water.add_s_alpha_beta('c_H_in_H2O')
# Define the materials file.
model.materials = (fuel, clad, hot_water)
# Instantiate ZCylinder surfaces
fuel_or = openmc.ZCylinder(x0=0, y0=0, R=0.39218, name='Fuel OR')
clad_or = openmc.ZCylinder(x0=0, y0=0, R=0.45720, name='Clad OR')
# Create boundary planes to surround the geometry
pitch = 21.42
min_x = openmc.XPlane(x0=-pitch/2, boundary_type='reflective')
max_x = openmc.XPlane(x0=+pitch/2, boundary_type='reflective')
min_y = openmc.YPlane(y0=-pitch/2, boundary_type='reflective')
max_y = openmc.YPlane(y0=+pitch/2, boundary_type='reflective')
# Create a fuel pin universe
fuel_pin_universe = openmc.Universe(name='Fuel Pin')
fuel_cell = openmc.Cell(name='fuel', fill=fuel, region=-fuel_or)
clad_cell = openmc.Cell(name='clad', fill=clad, region=+fuel_or & -clad_or)
hot_water_cell = openmc.Cell(name='hot water', fill=hot_water, region=+clad_or)
fuel_pin_universe.add_cells([fuel_cell, clad_cell, hot_water_cell])
# Create a control rod guide tube universe
guide_tube_universe = openmc.Universe(name='Guide Tube')
gt_inner_cell = openmc.Cell(name='guide tube inner water', fill=hot_water,
region=-fuel_or)
gt_clad_cell = openmc.Cell(name='guide tube clad', fill=clad,
region=+fuel_or & -clad_or)
gt_outer_cell = openmc.Cell(name='guide tube outer water', fill=hot_water,
region=+clad_or)
guide_tube_universe.add_cells([gt_inner_cell, gt_clad_cell, gt_outer_cell])
# Create fuel assembly Lattice
assembly = openmc.RectLattice(name='Fuel Assembly')
assembly.pitch = (pitch/17, pitch/17)
assembly.lower_left = (-pitch/2, -pitch/2)
# Create array indices for guide tube locations in lattice
template_x = np.array([5, 8, 11, 3, 13, 2, 5, 8, 11, 14, 2, 5, 8,
11, 14, 2, 5, 8, 11, 14, 3, 13, 5, 8, 11])
template_y = np.array([2, 2, 2, 3, 3, 5, 5, 5, 5, 5, 8, 8, 8, 8,
8, 11, 11, 11, 11, 11, 13, 13, 14, 14, 14])
# Create 17x17 array of universes
assembly.universes = np.tile(fuel_pin_universe, (17, 17))
assembly.universes[template_x, template_y] = guide_tube_universe
# Create root Cell
root_cell = openmc.Cell(name='root cell', fill=assembly)
root_cell.region = +min_x & -max_x & +min_y & -max_y
# Create root Universe
model.geometry.root_universe = openmc.Universe(name='root universe')
model.geometry.root_universe.add_cell(root_cell)
model.settings.batches = 10
model.settings.inactive = 5
model.settings.particles = 100
model.settings.source = openmc.Source(space=openmc.stats.Box(
[-pitch/2, -pitch/2, -1], [pitch/2, pitch/2, 1], only_fissionable=True))
plot = openmc.Plot()
plot.origin = (0.0, 0.0, 0)
plot.width = (21.42, 21.42)
plot.pixels = (300, 300)
plot.color_by = 'material'
model.plots.append(plot)
return model
def slab_mg(reps=None, as_macro=True):
"""Create a one-group, 1D slab model.
Parameters
----------
reps : list, optional
List of angular representations. Each item corresponds to materials and
dictates the angular representation of the multi-group cross
sections---isotropic ('iso') or angle-dependent ('ang'), and if Legendre
scattering or tabular scattering ('mu') is used. Thus, items can be
'ang', 'ang_mu', 'iso', or 'iso_mu'.
as_macro : bool, optional
Whether :class:`openmc.Macroscopic` is used
Returns
-------
model : openmc.model.Model
One-group, 1D slab model
"""
model = openmc.model.Model()
# Define materials needed for 1D/1G slab problem
mat_names = ['uo2', 'clad', 'lwtr']
mgxs_reps = ['ang', 'ang_mu', 'iso', 'iso_mu']
if reps is None:
reps = mgxs_reps
xs = []
i = 0
for mat in mat_names:
for rep in reps:
i += 1
if as_macro:
xs.append(openmc.Macroscopic(mat + '_' + rep))
m = openmc.Material(name=str(i))
m.set_density('macro', 1.)
m.add_macroscopic(xs[-1])
else:
xs.append(openmc.Nuclide(mat + '_' + rep))
m = openmc.Material(name=str(i))
m.set_density('atom/b-cm', 1.)
m.add_nuclide(xs[-1].name, 1.0, 'ao')
model.materials.append(m)
# Define the materials file
model.xs_data = xs
model.materials.cross_sections = "../1d_mgxs.h5"
# Define surfaces.
# Assembly/Problem Boundary
left = openmc.XPlane(x0=0.0, boundary_type='reflective')
right = openmc.XPlane(x0=10.0, boundary_type='reflective')
bottom = openmc.YPlane(y0=0.0, boundary_type='reflective')
top = openmc.YPlane(y0=10.0, boundary_type='reflective')
# for each material add a plane
planes = [openmc.ZPlane(z0=0.0, boundary_type='reflective')]
dz = round(5. / float(len(model.materials)), 4)
for i in range(len(model.materials) - 1):
planes.append(openmc.ZPlane(z0=dz * float(i + 1)))
planes.append(openmc.ZPlane(z0=5.0, boundary_type='reflective'))
# Define cells for each material
model.geometry.root_universe = openmc.Universe(name='root universe')
xy = +left & -right & +bottom & -top
for i, mat in enumerate(model.materials):
c = openmc.Cell(fill=mat, region=xy & +planes[i] & -planes[i + 1])
model.geometry.root_universe.add_cell(c)
model.settings.batches = 10
model.settings.inactive = 5
model.settings.particles = 100
model.settings.source = openmc.Source(space=openmc.stats.Box(
[0.0, 0.0, 0.0], [10.0, 10.0, 5.]))
model.settings.energy_mode = "multi-group"
plot = openmc.Plot()
plot.filename = 'mat'
plot.origin = (5.0, 5.0, 2.5)
plot.width = (2.5, 2.5)
plot.basis = 'xz'
plot.pixels = (3000, 3000)
plot.color_by = 'material'
model.plots.append(plot)
return model

9
openmc/geometry.py
View file

@ -1,4 +1,5 @@
from collections import OrderedDict, Iterable
from copy import deepcopy
from xml.etree import ElementTree as ET
from six import string_types
@ -497,3 +498,11 @@ class Geometry(object):
# Recursively traverse the CSG tree to count all cell instances
self.root_universe._determine_paths()
def clone(self):
"""Create a copy of this geometry with new unique IDs for all of its
enclosed materials, surfaces, cells, universes and lattices."""
clone = deepcopy(self)
clone.root_universe = self.root_universe.clone()
return clone

46
openmc/lattice.py
View file

@ -2,6 +2,7 @@ from __future__ import division
from abc import ABCMeta
from collections import OrderedDict, Iterable
from copy import deepcopy
from math import sqrt, floor
from numbers import Real, Integral
from xml.etree import ElementTree as ET
@ -414,6 +415,51 @@ class Lattice(object):
return []
return [(self, idx)] + u.find(p)
def clone(self, memo=None):
"""Create a copy of this lattice with a new unique ID, and clones
all universes within this lattice.
Parameters
----------
memo : dict or None
A nested dictionary of previously cloned objects. This parameter
is used internally and should not be specified by the user.
Returns
-------
clone : openmc.Lattice
The clone of this lattice
"""
if memo is None:
memo = {}
# If no nemoize'd clone exists, instantiate one
if self not in memo:
clone = deepcopy(self)
clone.id = None
if self.outer is not None:
clone.outer = self.outer.clone(memo)
# Assign universe clones to the lattice clone
for i in self.indices:
if isinstance(self, RectLattice):
clone.universes[i] = self.universes[i].clone(memo)
else:
if self.ndim == 2:
clone.universes[i[0]][i[1]] = \
self.universes[i[0]][i[1]].clone(memo)
else:
clone.universes[i[0]][i[1]][i[2]] = \
self.universes[i[0]][i[1]][i[2]].clone(memo)
# Memoize the clone
memo[self] = clone
return memo[self]
class RectLattice(Lattice):
"""A lattice consisting of rectangular prisms.

29
openmc/material.py
View file

@ -799,6 +799,35 @@ class Material(object):
return nuclides
def clone(self, memo=None):
"""Create a copy of this material with a new unique ID.
Parameters
----------
memo : dict or None
A nested dictionary of previously cloned objects. This parameter
is used internally and should not be specified by the user.
Returns
-------
clone : openmc.Material
The clone of this material
"""
if memo is None:
memo = {}
# If no nemoize'd clone exists, instantiate one
if self not in memo:
clone = deepcopy(self)
clone.id = None
# Memoize the clone
memo[self] = clone
return memo[self]
def _get_nuclide_xml(self, nuclide, distrib=False):
xml_element = ET.Element("nuclide")
xml_element.set("name", nuclide[0].name)

141
openmc/model/model.py
View file

@ -1,26 +1,34 @@
from collections import Iterable
import openmc
from openmc.checkvalue import check_type
class Model(object):
"""OpenMC model container for the openmc.Geometry, openmc.Materials,
openmc.Settings, openmc.Tallies, openmc.CMFD objects, and openmc.Plot
objects
"""Model container.
This class can be used to store instances of :class:`openmc.Geometry`,
:class:`openmc.Materials`, :class:`openmc.Settings`,
:class:`openmc.Tallies`, :class:`openmc.Plots`, and :class:`openmc.CMFD`,
thus making a complete model. The :meth:`Model.export_to_xml` method will
export XML files for all attributes that have been set. If the
:meth:`Model.materials` attribute is not set, it will attempt to create a
``materials.xml`` file based on all materials appearing in the geometry.
Parameters
----------
geometry : openmc.Geometry
geometry : openmc.Geometry, optional
Geometry information
materials : openmc.Materials
materials : openmc.Materials, optional
Materials information
settings : openmc.Settings
settings : openmc.Settings, optional
Settings information
tallies : openmc.Tallies
Tallies information, optional
cmfd : openmc.CMFD
CMFD information, optional
plots : openmc.Plots
Plot information, optional
tallies : openmc.Tallies, optional
Tallies information
cmfd : openmc.CMFD, optional
CMFD information
plots : openmc.Plots, optional
Plot information
Attributes
----------
@ -39,25 +47,25 @@ class Model(object):
"""
def __init__(self, geometry, materials, settings, tallies=None, cmfd=None,
plots=None):
self.geometry = geometry
self.materials = materials
self.settings = settings
if tallies:
self.tallies = tallies
else:
self._tallies = openmc.Tallies()
if cmfd:
self.cmfd = cmfd
else:
self._cmfd = None
if plots:
self.plots = plots
else:
self.plots = openmc.Plots()
def __init__(self, geometry=None, materials=None, settings=None,
tallies=None, cmfd=None, plots=None):
self.geometry = openmc.Geometry()
self.materials = openmc.Materials()
self.settings = openmc.Settings()
self.cmfd = cmfd
self.tallies = openmc.Tallies()
self.plots = openmc.Plots()
self.sp = None
if geometry is not None:
self.geometry = geometry
if materials is not None:
self.materials = materials
if settings is not None:
self.settings = settings
if tallies is not None:
self.tallies = tallies
if plots is not None:
self.plots = plots
@property
def geometry(self):
@ -90,8 +98,13 @@ class Model(object):
@materials.setter
def materials(self, materials):
check_type('materials', materials, openmc.Materials)
self._materials = materials
check_type('materials', materials, Iterable, openmc.Material)
if isinstance(materials, openmc.Materials):
self._materials = materials
else:
del self._materials[:]
for mat in materials:
self._materials.append(mat)
@settings.setter
def settings(self, settings):
@ -100,33 +113,55 @@ class Model(object):
@tallies.setter
def tallies(self, tallies):
check_type('tallies', tallies, openmc.Tallies)
self._tallies = tallies
check_type('tallies', tallies, Iterable, openmc.Tally)
if isinstance(tallies, openmc.Tallies):
self._tallies = tallies
else:
del self._tallies[:]
for tally in tallies:
self._tallies.append(tally)
@cmfd.setter
def cmfd(self, cmfd):
check_type('cmfd', cmfd, openmc.CMFD)
check_type('cmfd', cmfd, (openmc.CMFD, type(None)))
self._cmfd = cmfd
@plots.setter
def plots(self, plots):
check_type('plots', plots, openmc.Plots)
self._plots = plots
check_type('plots', plots, Iterable, openmc.Plot)
if isinstance(plots, openmc.Plots):
self._plots = plots
else:
del self._plots[:]
for plot in plots:
self._plots.append(plot)
def export_to_xml(self):
"""Export model settings to XML files.
"""Export model to XML files.
"""
self.geometry.export_to_xml()
self.materials.export_to_xml()
self.settings.export_to_xml()
self.tallies.export_to_xml()
self.geometry.export_to_xml()
# If a materials collection was specified, export it. Otherwise, look
# for all materials in the geometry and use that to automatically build
# a collection.
if self.materials:
self.materials.export_to_xml()
else:
materials = openmc.Materials(self.geometry.get_all_materials()
.values())
materials.export_to_xml()
if self.tallies:
self.tallies.export_to_xml()
if self.cmfd is not None:
self.cmfd.export_to_xml()
self.plots.export_to_xml()
if self.plots:
self.plots.export_to_xml()
def run(self, **kwargs):
"""Creates the XML files, runs OpenMC, and loads the statepoint.
"""Creates the XML files, runs OpenMC, and returns k-effective
Parameters
----------
@ -136,29 +171,19 @@ class Model(object):
Returns
-------
2-tuple of float
k_combined from the statepoint
Combined estimator of k-effective from the statepoint
"""
self.export_to_xml()
return_code = openmc.run(**kwargs)
assert (return_code == 0), "OpenMC did not execute successfully"
statepoint_batches = self.settings.batches
n = self.settings.batches
if self.settings.statepoint is not None:
if 'batches' in self.settings.statepoint:
statepoint_batches = self.settings.statepoint['batches'][-1]
self.sp = \
openmc.StatePoint('statepoint.{}.h5'.format(statepoint_batches))
n = self.settings.statepoint['batches'][-1]
return self.sp.k_combined
def close(self):
"""Close the statepoint and summary files
"""
if self.sp is not None:
self.sp._f.close()
self.sp.summary._f.close()
with openmc.StatePoint('statepoint.{}.h5'.format(n)) as sp:
return sp.k_combined

98
openmc/region.py
View file

@ -1,5 +1,6 @@
from abc import ABCMeta, abstractmethod
from collections import Iterable, OrderedDict
from copy import deepcopy
from six import add_metaclass
import numpy as np
@ -221,6 +222,31 @@ class Region(object):
# at the end
return output[0]
@abstractmethod
def clone(self, memo=None):
"""Create a copy of this region - each of the surfaces in the
region's nodes will be cloned and will have new unique IDs.
Parameters
----------
memo : dict or None
A nested dictionary of previously cloned objects. This parameter
is used internally and should not be specified by the user.
Returns
-------
clone : openmc.Region
The clone of this region
Raises
------
NotImplementedError
This method is not implemented for the abstract region class.
"""
raise NotImplementedError('The clone method is not implemented for '
'the abstract region class.')
class Intersection(Region):
r"""Intersection of two or more regions.
@ -300,6 +326,30 @@ class Intersection(Region):
check_type('nodes', nodes, Iterable, Region)
self._nodes = nodes
def clone(self, memo=None):
"""Create a copy of this region - each of the surfaces in the
intersection's nodes will be cloned and will have new unique IDs.
Parameters
----------
memo : dict or None
A nested dictionary of previously cloned objects. This parameter
is used internally and should not be specified by the user.
Returns
-------
clone : openmc.Intersection
The clone of this intersection
"""
if memo is None:
memo = {}
clone = deepcopy(self)
clone.nodes = [n.clone(memo) for n in self.nodes]
return clone
class Union(Region):
r"""Union of two or more regions.
@ -377,6 +427,30 @@ class Union(Region):
check_type('nodes', nodes, Iterable, Region)
self._nodes = nodes
def clone(self, memo=None):
"""Create a copy of this region - each of the surfaces in the
union's nodes will be cloned and will have new unique IDs.
Parameters
----------
memo : dict or None
A nested dictionary of previously cloned objects. This parameter
is used internally and should not be specified by the user.
Returns
-------
clone : openmc.Union
The clone of this union
"""
if memo is None:
memo = {}
clone = copy.deepcopy(self)
clone.nodes = [n.clone(memo) for n in self.nodes]
return clone
class Complement(Region):
"""Complement of a region.
@ -473,3 +547,27 @@ class Complement(Region):
for region in self.node:
surfaces = region.get_surfaces(surfaces)
return surfaces
def clone(self, memo=None):
"""Create a copy of this region - each of the surfaces in the
complement's node will be cloned and will have new unique IDs.
Parameters
----------
memo : dict or None
A nested dictionary of previously cloned objects. This parameter
is used internally and should not be specified by the user.
Returns
-------
clone : openmc.Complement
The clone of this complement
"""
if memo is None:
memo = {}
clone = copy.deepcopy(self)
clone.node = self.node.clone(memo)
return clone

4
openmc/search.py
View file

@ -51,10 +51,6 @@ def _search_keff(guess, target, model_builder, model_args, print_iterations,
# Run the model and obtain keff
keff = model.run(output=print_output)
# Close the model to ensure HDF5 will allow access during the next
# OpenMC execution
model.close()
# Record the history
guesses.append(guess)
results.append(keff)

6
openmc/statepoint.py
View file

@ -138,6 +138,12 @@ class StatePoint(object):
vol = openmc.VolumeCalculation.from_hdf5(path_i)
self.add_volume_information(vol)
def __enter__(self):
return self
def __exit__(self, *exc):
self._f.close()
@property
def cmfd_on(self):
return self._f.attrs['cmfd_on'] > 0

59
openmc/surface.py
View file

@ -1,5 +1,6 @@
from abc import ABCMeta
from collections import Iterable, OrderedDict
from copy import deepcopy
from numbers import Real, Integral
from xml.etree import ElementTree as ET
from math import sqrt
@ -82,6 +83,9 @@ class Surface(object):
def __pos__(self):
return Halfspace(self, '+')
def __hash__(self):
return hash(repr(self))
def __repr__(self):
string = 'Surface\n'
string += '{0: <16}{1}{2}\n'.format('\tID', '=\t', self._id)
@ -171,6 +175,35 @@ class Surface(object):
return (np.array([-np.inf, -np.inf, -np.inf]),
np.array([np.inf, np.inf, np.inf]))
def clone(self, memo=None):
"""Create a copy of this surface with a new unique ID.
Parameters
----------
memo : dict or None
A nested dictionary of previously cloned objects. This parameter
is used internally and should not be specified by the user.
Returns
-------
clone : openmc.Surface
The clone of this surface
"""
if memo is None:
memo = {}
# If no nemoize'd clone exists, instantiate one
if self not in memo:
clone = deepcopy(self)
clone.id = None
# Memoize the clone
memo[self] = clone
return memo[self]
def to_xml_element(self):
"""Return XML representation of the surface
@ -1821,7 +1854,7 @@ class Halfspace(Region):
def __init__(self, surface, side):
self.surface = surface
self.side = side
def __and__(self, other):
if isinstance(other, Intersection):
return Intersection(self, *other.nodes)
@ -1902,6 +1935,30 @@ class Halfspace(Region):
surfaces[self.surface.id] = self.surface
return surfaces
def clone(self, memo=None):
"""Create a copy of this halfspace, with a cloned surface with a
unique ID.
Parameters
----------
memo : dict or None
A nested dictionary of previously cloned objects. This parameter
is used internally and should not be specified by the user.
Returns
-------
clone : openmc.Halfspace
The clone of this halfspace
"""
if memo is None:
memo = dict
clone = deepcopy(self)
clone.surface = self.surface.clone(memo)
return clone
def get_rectangular_prism(width, height, axis='z', origin=(0., 0.),
boundary_type='transmission'):

37
openmc/universe.py
View file

@ -1,6 +1,6 @@
from __future__ import division
from copy import copy
from collections import OrderedDict, Iterable
from copy import copy, deepcopy
from numbers import Integral, Real
import random
import sys
@ -517,6 +517,41 @@ class Universe(object):
return universes
def clone(self, memo=None):
"""Create a copy of this universe with a new unique ID, and clones
all cells within this universe.
Parameters
----------
memo : dict or None
A nested dictionary of previously cloned objects. This parameter
is used internally and should not be specified by the user.
Returns
-------
clone : openmc.Universe
The clone of this universe
"""
if memo is None:
memo = {}
# If no nemoize'd clone exists, instantiate one
if self not in memo:
clone = deepcopy(self)
clone.id = None
# Clone all cells for the universe clone
clone._cells = OrderedDict()
for cell in self._cells.values():
clone.add_cell(cell.clone(memo))
# Memoize the clone
memo[self] = clone
return memo[self]
def create_xml_subelement(self, xml_element):
# Iterate over all Cells
for cell_id, cell in self._cells.items():

4
scripts/openmc-track-to-vtk
View file

@ -42,8 +42,8 @@ def main():
for fname in args.input:
# Write coordinate values to points array.
track = h5py.File(fname)
n_particles = track['n_particles'].value
n_coords = track['n_coords']
n_particles = track.attrs['n_particles']
n_coords = track.attrs['n_coords']
coords = []
for i in range(n_particles):
coords.append(track['coordinates_' + str(i + 1)].value)

7
src/input_xml.F90
View file

@ -5303,6 +5303,13 @@ contains
do i = 1, size(res_scat_nuclides)
if (nuc % name == res_scat_nuclides(i)) then
! Make sure nuclide has 0K data
if (.not. allocated(nuc % energy_0K)) then
call fatal_error("Cannot treat " // trim(nuc % name) // " as a &
&resonant scatterer because 0 K elastic scattering data is not &
&present.")
end if
! Set nuclide to be resonant
nuc % resonant = .true.

798
tests/input_set.py
View file

@ -1,798 +0,0 @@
import numpy as np
import openmc
from openmc.source import Source
from openmc.stats import Box
class InputSet(object):
def __init__(self):
self.settings = openmc.Settings()
self.materials = openmc.Materials()
self.geometry = openmc.Geometry()
self.tallies = None
self.plots = None
def export(self):
self.settings.export_to_xml()
self.materials.export_to_xml()
self.geometry.export_to_xml()
if self.tallies is not None:
self.tallies.export_to_xml()
if self.plots is not None:
self.plots.export_to_xml()
def build_default_materials_and_geometry(self):
# Define materials.
fuel = openmc.Material(name='Fuel', material_id=1)
fuel.set_density('g/cm3', 10.062)
fuel.add_nuclide("U234", 4.9476e-6)
fuel.add_nuclide("U235", 4.8218e-4)
fuel.add_nuclide("U238", 2.1504e-2)
fuel.add_nuclide("Xe135", 1.0801e-8)
fuel.add_nuclide("O16", 4.5737e-2)
clad = openmc.Material(name='Cladding', material_id=2)
clad.set_density('g/cm3', 5.77)
clad.add_nuclide("Zr90", 0.5145)
clad.add_nuclide("Zr91", 0.1122)
clad.add_nuclide("Zr92", 0.1715)
clad.add_nuclide("Zr94", 0.1738)
clad.add_nuclide("Zr96", 0.0280)
cold_water = openmc.Material(name='Cold borated water', material_id=3)
cold_water.set_density('atom/b-cm', 0.07416)
cold_water.add_nuclide("H1", 2.0)
cold_water.add_nuclide("O16", 1.0)
cold_water.add_nuclide("B10", 6.490e-4)
cold_water.add_nuclide("B11", 2.689e-3)
cold_water.add_s_alpha_beta('c_H_in_H2O')
hot_water = openmc.Material(name='Hot borated water', material_id=4)
hot_water.set_density('atom/b-cm', 0.06614)
hot_water.add_nuclide("H1", 2.0)
hot_water.add_nuclide("O16", 1.0)
hot_water.add_nuclide("B10", 6.490e-4)
hot_water.add_nuclide("B11", 2.689e-3)
hot_water.add_s_alpha_beta('c_H_in_H2O')
rpv_steel = openmc.Material(name='Reactor pressure vessel steel',
material_id=5)
rpv_steel.set_density('g/cm3', 7.9)
rpv_steel.add_nuclide("Fe54", 0.05437098, 'wo')
rpv_steel.add_nuclide("Fe56", 0.88500663, 'wo')
rpv_steel.add_nuclide("Fe57", 0.0208008, 'wo')
rpv_steel.add_nuclide("Fe58", 0.00282159, 'wo')
rpv_steel.add_nuclide("Ni58", 0.0067198, 'wo')
rpv_steel.add_nuclide("Ni60", 0.0026776, 'wo')
rpv_steel.add_nuclide("Mn55", 0.01, 'wo')
rpv_steel.add_nuclide("Cr52", 0.002092475, 'wo')
rpv_steel.add_nuclide("C0", 0.0025, 'wo')
rpv_steel.add_nuclide("Cu63", 0.0013696, 'wo')
lower_rad_ref = openmc.Material(name='Lower radial reflector',
material_id=6)
lower_rad_ref.set_density('g/cm3', 4.32)
lower_rad_ref.add_nuclide("H1", 0.0095661, 'wo')
lower_rad_ref.add_nuclide("O16", 0.0759107, 'wo')
lower_rad_ref.add_nuclide("B10", 3.08409e-5, 'wo')
lower_rad_ref.add_nuclide("B11", 1.40499e-4, 'wo')
lower_rad_ref.add_nuclide("Fe54", 0.035620772088, 'wo')
lower_rad_ref.add_nuclide("Fe56", 0.579805982228, 'wo')
lower_rad_ref.add_nuclide("Fe57", 0.01362750048, 'wo')
lower_rad_ref.add_nuclide("Fe58", 0.001848545204, 'wo')
lower_rad_ref.add_nuclide("Ni58", 0.055298376566, 'wo')
lower_rad_ref.add_nuclide("Mn55", 0.0182870, 'wo')
lower_rad_ref.add_nuclide("Cr52", 0.145407678031, 'wo')
lower_rad_ref.add_s_alpha_beta('c_H_in_H2O')
upper_rad_ref = openmc.Material(name='Upper radial reflector /'
'Top plate region', material_id=7)
upper_rad_ref.set_density('g/cm3', 4.28)
upper_rad_ref.add_nuclide("H1", 0.0086117, 'wo')
upper_rad_ref.add_nuclide("O16", 0.0683369, 'wo')
upper_rad_ref.add_nuclide("B10", 2.77638e-5, 'wo')
upper_rad_ref.add_nuclide("B11", 1.26481e-4, 'wo')
upper_rad_ref.add_nuclide("Fe54", 0.035953677186, 'wo')
upper_rad_ref.add_nuclide("Fe56", 0.585224740891, 'wo')
upper_rad_ref.add_nuclide("Fe57", 0.01375486056, 'wo')
upper_rad_ref.add_nuclide("Fe58", 0.001865821363, 'wo')
upper_rad_ref.add_nuclide("Ni58", 0.055815129186, 'wo')
upper_rad_ref.add_nuclide("Mn55", 0.0184579, 'wo')
upper_rad_ref.add_nuclide("Cr52", 0.146766614995, 'wo')
upper_rad_ref.add_s_alpha_beta('c_H_in_H2O')
bot_plate = openmc.Material(name='Bottom plate region', material_id=8)
bot_plate.set_density('g/cm3', 7.184)
bot_plate.add_nuclide("H1", 0.0011505, 'wo')
bot_plate.add_nuclide("O16", 0.0091296, 'wo')
bot_plate.add_nuclide("B10", 3.70915e-6, 'wo')
bot_plate.add_nuclide("B11", 1.68974e-5, 'wo')
bot_plate.add_nuclide("Fe54", 0.03855611055, 'wo')
bot_plate.add_nuclide("Fe56", 0.627585036425, 'wo')
bot_plate.add_nuclide("Fe57", 0.014750478, 'wo')
bot_plate.add_nuclide("Fe58", 0.002000875025, 'wo')
bot_plate.add_nuclide("Ni58", 0.059855207342, 'wo')
bot_plate.add_nuclide("Mn55", 0.0197940, 'wo')
bot_plate.add_nuclide("Cr52", 0.157390026871, 'wo')
bot_plate.add_s_alpha_beta('c_H_in_H2O')
bot_nozzle = openmc.Material(name='Bottom nozzle region',
material_id=9)
bot_nozzle.set_density('g/cm3', 2.53)
bot_nozzle.add_nuclide("H1", 0.0245014, 'wo')
bot_nozzle.add_nuclide("O16", 0.1944274, 'wo')
bot_nozzle.add_nuclide("B10", 7.89917e-5, 'wo')
bot_nozzle.add_nuclide("B11", 3.59854e-4, 'wo')
bot_nozzle.add_nuclide("Fe54", 0.030411411144, 'wo')
bot_nozzle.add_nuclide("Fe56", 0.495012237964, 'wo')
bot_nozzle.add_nuclide("Fe57", 0.01163454624, 'wo')
bot_nozzle.add_nuclide("Fe58", 0.001578204652, 'wo')
bot_nozzle.add_nuclide("Ni58", 0.047211231662, 'wo')
bot_nozzle.add_nuclide("Mn55", 0.0156126, 'wo')
bot_nozzle.add_nuclide("Cr52", 0.124142524198, 'wo')
bot_nozzle.add_s_alpha_beta('c_H_in_H2O')
top_nozzle = openmc.Material(name='Top nozzle region', material_id=10)
top_nozzle.set_density('g/cm3', 1.746)
top_nozzle.add_nuclide("H1", 0.0358870, 'wo')
top_nozzle.add_nuclide("O16", 0.2847761, 'wo')
top_nozzle.add_nuclide("B10", 1.15699e-4, 'wo')
top_nozzle.add_nuclide("B11", 5.27075e-4, 'wo')
top_nozzle.add_nuclide("Fe54", 0.02644016154, 'wo')
top_nozzle.add_nuclide("Fe56", 0.43037146399, 'wo')
top_nozzle.add_nuclide("Fe57", 0.0101152584, 'wo')
top_nozzle.add_nuclide("Fe58", 0.00137211607, 'wo')
top_nozzle.add_nuclide("Ni58", 0.04104621835, 'wo')
top_nozzle.add_nuclide("Mn55", 0.0135739, 'wo')
top_nozzle.add_nuclide("Cr52", 0.107931450781, 'wo')
top_nozzle.add_s_alpha_beta('c_H_in_H2O')
top_fa = openmc.Material(name='Top of fuel assemblies', material_id=11)
top_fa.set_density('g/cm3', 3.044)
top_fa.add_nuclide("H1", 0.0162913, 'wo')
top_fa.add_nuclide("O16", 0.1292776, 'wo')
top_fa.add_nuclide("B10", 5.25228e-5, 'wo')
top_fa.add_nuclide("B11", 2.39272e-4, 'wo')
top_fa.add_nuclide("Zr90", 0.43313403903, 'wo')
top_fa.add_nuclide("Zr91", 0.09549277374, 'wo')
top_fa.add_nuclide("Zr92", 0.14759527104, 'wo')
top_fa.add_nuclide("Zr94", 0.15280552077, 'wo')
top_fa.add_nuclide("Zr96", 0.02511169542, 'wo')
top_fa.add_s_alpha_beta('c_H_in_H2O')
bot_fa = openmc.Material(name='Bottom of fuel assemblies',
material_id=12)
bot_fa.set_density('g/cm3', 1.762)
bot_fa.add_nuclide("H1", 0.0292856, 'wo')
bot_fa.add_nuclide("O16", 0.2323919, 'wo')
bot_fa.add_nuclide("B10", 9.44159e-5, 'wo')
bot_fa.add_nuclide("B11", 4.30120e-4, 'wo')
bot_fa.add_nuclide("Zr90", 0.3741373658, 'wo')
bot_fa.add_nuclide("Zr91", 0.0824858164, 'wo')
bot_fa.add_nuclide("Zr92", 0.1274914944, 'wo')
bot_fa.add_nuclide("Zr94", 0.1319920622, 'wo')
bot_fa.add_nuclide("Zr96", 0.0216912612, 'wo')
bot_fa.add_s_alpha_beta('c_H_in_H2O')
# Define the materials file.
self.materials += (fuel, clad, cold_water, hot_water, rpv_steel,
lower_rad_ref, upper_rad_ref, bot_plate,
bot_nozzle, top_nozzle, top_fa, bot_fa)
# Define surfaces.
s1 = openmc.ZCylinder(R=0.41, surface_id=1)
s2 = openmc.ZCylinder(R=0.475, surface_id=2)
s3 = openmc.ZCylinder(R=0.56, surface_id=3)
s4 = openmc.ZCylinder(R=0.62, surface_id=4)
s5 = openmc.ZCylinder(R=187.6, surface_id=5)
s6 = openmc.ZCylinder(R=209.0, surface_id=6)
s7 = openmc.ZCylinder(R=229.0, surface_id=7)
s8 = openmc.ZCylinder(R=249.0, surface_id=8)
s8.boundary_type = 'vacuum'
s31 = openmc.ZPlane(z0=-229.0, surface_id=31)
s31.boundary_type = 'vacuum'
s32 = openmc.ZPlane(z0=-199.0, surface_id=32)
s33 = openmc.ZPlane(z0=-193.0, surface_id=33)
s34 = openmc.ZPlane(z0=-183.0, surface_id=34)
s35 = openmc.ZPlane(z0=0.0, surface_id=35)
s36 = openmc.ZPlane(z0=183.0, surface_id=36)
s37 = openmc.ZPlane(z0=203.0, surface_id=37)
s38 = openmc.ZPlane(z0=215.0, surface_id=38)
s39 = openmc.ZPlane(z0=223.0, surface_id=39)
s39.boundary_type = 'vacuum'
# Define pin cells.
fuel_cold = openmc.Universe(name='Fuel pin, cladding, cold water',
universe_id=1)
c21 = openmc.Cell(cell_id=21)
c21.region = -s1
c21.fill = fuel
c22 = openmc.Cell(cell_id=22)
c22.region = +s1 & -s2
c22.fill = clad
c23 = openmc.Cell(cell_id=23)
c23.region = +s2
c23.fill = cold_water
fuel_cold.add_cells((c21, c22, c23))
tube_cold = openmc.Universe(name='Instrumentation guide tube, '
'cold water', universe_id=2)
c24 = openmc.Cell(cell_id=24)
c24.region = -s3
c24.fill = cold_water
c25 = openmc.Cell(cell_id=25)
c25.region = +s3 & -s4
c25.fill = clad
c26 = openmc.Cell(cell_id=26)
c26.region = +s4
c26.fill = cold_water
tube_cold.add_cells((c24, c25, c26))
fuel_hot = openmc.Universe(name='Fuel pin, cladding, hot water',
universe_id=3)
c27 = openmc.Cell(cell_id=27)
c27.region = -s1
c27.fill = fuel
c28 = openmc.Cell(cell_id=28)
c28.region = +s1 & -s2
c28.fill = clad
c29 = openmc.Cell(cell_id=29)
c29.region = +s2
c29.fill = hot_water
fuel_hot.add_cells((c27, c28, c29))
tube_hot = openmc.Universe(name='Instrumentation guide tube, hot water',
universe_id=4)
c30 = openmc.Cell(cell_id=30)
c30.region = -s3
c30.fill = hot_water
c31 = openmc.Cell(cell_id=31)
c31.region = +s3 & -s4
c31.fill = clad
c32 = openmc.Cell(cell_id=32)
c32.region = +s4
c32.fill = hot_water
tube_hot.add_cells((c30, c31, c32))
# Define fuel lattices.
l100 = openmc.RectLattice(name='Fuel assembly (lower half)',
lattice_id=100)
l100.lower_left = (-10.71, -10.71)
l100.pitch = (1.26, 1.26)
l100.universes = [
[fuel_cold]*17,
[fuel_cold]*17,
[fuel_cold]*5 + [tube_cold] + [fuel_cold]*2 + [tube_cold]
+ [fuel_cold]*2 + [tube_cold] + [fuel_cold]*5,
[fuel_cold]*3 + [tube_cold] + [fuel_cold]*9 + [tube_cold]
+ [fuel_cold]*3,
[fuel_cold]*17,
[fuel_cold]*2 + [tube_cold] + [fuel_cold]*2 + [tube_cold]
+ [fuel_cold]*2 + [tube_cold] + [fuel_cold]*2 + [tube_cold]
+ [fuel_cold]*2 + [tube_cold] + [fuel_cold]*2,
[fuel_cold]*17,
[fuel_cold]*17,
[fuel_cold]*2 + [tube_cold] + [fuel_cold]*2 + [tube_cold]
+ [fuel_cold]*2 + [tube_cold] + [fuel_cold]*2 + [tube_cold]
+ [fuel_cold]*2 + [tube_cold] + [fuel_cold]*2,
[fuel_cold]*17,
[fuel_cold]*17,
[fuel_cold]*2 + [tube_cold] + [fuel_cold]*2 + [tube_cold]
+ [fuel_cold]*2 + [tube_cold] + [fuel_cold]*2 + [tube_cold]
+ [fuel_cold]*2 + [tube_cold] + [fuel_cold]*2,
[fuel_cold]*17,
[fuel_cold]*3 + [tube_cold] + [fuel_cold]*9 + [tube_cold]
+ [fuel_cold]*3,
[fuel_cold]*5 + [tube_cold] + [fuel_cold]*2 + [tube_cold]
+ [fuel_cold]*2 + [tube_cold] + [fuel_cold]*5,
[fuel_cold]*17,
[fuel_cold]*17 ]
l101 = openmc.RectLattice(name='Fuel assembly (upper half)',
lattice_id=101)
l101.lower_left = (-10.71, -10.71)
l101.pitch = (1.26, 1.26)
l101.universes = [
[fuel_hot]*17,
[fuel_hot]*17,
[fuel_hot]*5 + [tube_hot] + [fuel_hot]*2 + [tube_hot]
+ [fuel_hot]*2 + [tube_hot] + [fuel_hot]*5,
[fuel_hot]*3 + [tube_hot] + [fuel_hot]*9 + [tube_hot]
+ [fuel_hot]*3,
[fuel_hot]*17,
[fuel_hot]*2 + [tube_hot] + [fuel_hot]*2 + [tube_hot]
+ [fuel_hot]*2 + [tube_hot] + [fuel_hot]*2 + [tube_hot]
+ [fuel_hot]*2 + [tube_hot] + [fuel_hot]*2,
[fuel_hot]*17,
[fuel_hot]*17,
[fuel_hot]*2 + [tube_hot] + [fuel_hot]*2 + [tube_hot]
+ [fuel_hot]*2 + [tube_hot] + [fuel_hot]*2 + [tube_hot]
+ [fuel_hot]*2 + [tube_hot] + [fuel_hot]*2,
[fuel_hot]*17,
[fuel_hot]*17,
[fuel_hot]*2 + [tube_hot] + [fuel_hot]*2 + [tube_hot]
+ [fuel_hot]*2 + [tube_hot] + [fuel_hot]*2 + [tube_hot]
+ [fuel_hot]*2 + [tube_hot] + [fuel_hot]*2,
[fuel_hot]*17,
[fuel_hot]*3 + [tube_hot] + [fuel_hot]*9 + [tube_hot]
+ [fuel_hot]*3,
[fuel_hot]*5 + [tube_hot] + [fuel_hot]*2 + [tube_hot]
+ [fuel_hot]*2 + [tube_hot] + [fuel_hot]*5,
[fuel_hot]*17,
[fuel_hot]*17 ]
# Define assemblies.
fa_cw = openmc.Universe(name='Water assembly (cold)', universe_id=5)
c50 = openmc.Cell(cell_id=50)
c50.region = +s34 & -s35
c50.fill = cold_water
fa_cw.add_cells((c50, ))
fa_hw = openmc.Universe(name='Water assembly (hot)', universe_id=7)
c70 = openmc.Cell(cell_id=70)
c70.region = +s35 & -s36
c70.fill = hot_water
fa_hw.add_cells((c70, ))
fa_cold = openmc.Universe(name='Fuel assembly (cold)', universe_id=6)
c60 = openmc.Cell(cell_id=60)
c60.region = +s34 & -s35
c60.fill = l100
fa_cold.add_cells((c60, ))
fa_hot = openmc.Universe(name='Fuel assembly (hot)', universe_id=8)
c80 = openmc.Cell(cell_id=80)
c80.region = +s35 & -s36
c80.fill = l101
fa_hot.add_cells((c80, ))
# Define core lattices
l200 = openmc.RectLattice(name='Core lattice (lower half)',
lattice_id=200)
l200.lower_left = (-224.91, -224.91)
l200.pitch = (21.42, 21.42)
l200.universes = [
[fa_cw]*21,
[fa_cw]*21,
[fa_cw]*7 + [fa_cold]*7 + [fa_cw]*7,
[fa_cw]*5 + [fa_cold]*11 + [fa_cw]*5,
[fa_cw]*4 + [fa_cold]*13 + [fa_cw]*4,
[fa_cw]*3 + [fa_cold]*15 + [fa_cw]*3,
[fa_cw]*3 + [fa_cold]*15 + [fa_cw]*3,
[fa_cw]*2 + [fa_cold]*17 + [fa_cw]*2,
[fa_cw]*2 + [fa_cold]*17 + [fa_cw]*2,
[fa_cw]*2 + [fa_cold]*17 + [fa_cw]*2,
[fa_cw]*2 + [fa_cold]*17 + [fa_cw]*2,
[fa_cw]*2 + [fa_cold]*17 + [fa_cw]*2,
[fa_cw]*2 + [fa_cold]*17 + [fa_cw]*2,
[fa_cw]*2 + [fa_cold]*17 + [fa_cw]*2,
[fa_cw]*3 + [fa_cold]*15 + [fa_cw]*3,
[fa_cw]*3 + [fa_cold]*15 + [fa_cw]*3,
[fa_cw]*4 + [fa_cold]*13 + [fa_cw]*4,
[fa_cw]*5 + [fa_cold]*11 + [fa_cw]*5,
[fa_cw]*7 + [fa_cold]*7 + [fa_cw]*7,
[fa_cw]*21,
[fa_cw]*21]
l201 = openmc.RectLattice(name='Core lattice (lower half)',
lattice_id=201)
l201.lower_left = (-224.91, -224.91)
l201.pitch = (21.42, 21.42)
l201.universes = [
[fa_hw]*21,
[fa_hw]*21,
[fa_hw]*7 + [fa_hot]*7 + [fa_hw]*7,
[fa_hw]*5 + [fa_hot]*11 + [fa_hw]*5,
[fa_hw]*4 + [fa_hot]*13 + [fa_hw]*4,
[fa_hw]*3 + [fa_hot]*15 + [fa_hw]*3,
[fa_hw]*3 + [fa_hot]*15 + [fa_hw]*3,
[fa_hw]*2 + [fa_hot]*17 + [fa_hw]*2,
[fa_hw]*2 + [fa_hot]*17 + [fa_hw]*2,
[fa_hw]*2 + [fa_hot]*17 + [fa_hw]*2,
[fa_hw]*2 + [fa_hot]*17 + [fa_hw]*2,
[fa_hw]*2 + [fa_hot]*17 + [fa_hw]*2,
[fa_hw]*2 + [fa_hot]*17 + [fa_hw]*2,
[fa_hw]*2 + [fa_hot]*17 + [fa_hw]*2,
[fa_hw]*3 + [fa_hot]*15 + [fa_hw]*3,
[fa_hw]*3 + [fa_hot]*15 + [fa_hw]*3,
[fa_hw]*4 + [fa_hot]*13 + [fa_hw]*4,
[fa_hw]*5 + [fa_hot]*11 + [fa_hw]*5,
[fa_hw]*7 + [fa_hot]*7 + [fa_hw]*7,
[fa_hw]*21,
[fa_hw]*21]
# Define root universe.
root = openmc.Universe(universe_id=0, name='root universe')
c1 = openmc.Cell(cell_id=1)
c1.region = -s6 & +s34 & -s35
c1.fill = l200
c2 = openmc.Cell(cell_id=2)
c2.region = -s6 & +s35 & -s36
c2.fill = l201
c3 = openmc.Cell(cell_id=3)
c3.region = -s7 & +s31 & -s32
c3.fill = bot_plate
c4 = openmc.Cell(cell_id=4)
c4.region = -s5 & +s32 & -s33
c4.fill = bot_nozzle
c5 = openmc.Cell(cell_id=5)
c5.region = -s5 & +s33 & -s34
c5.fill = bot_fa
c6 = openmc.Cell(cell_id=6)
c6.region = -s5 & +s36 & -s37
c6.fill = top_fa
c7 = openmc.Cell(cell_id=7)
c7.region = -s5 & +s37 & -s38
c7.fill = top_nozzle
c8 = openmc.Cell(cell_id=8)
c8.region = -s7 & +s38 & -s39
c8.fill = upper_rad_ref
c9 = openmc.Cell(cell_id=9)
c9.region = +s6 & -s7 & +s32 & -s38
c9.fill = bot_nozzle
c10 = openmc.Cell(cell_id=10)
c10.region = +s7 & -s8 & +s31 & -s39
c10.fill = rpv_steel
c11 = openmc.Cell(cell_id=11)
c11.region = +s5 & -s6 & +s32 & -s34
c11.fill = lower_rad_ref
c12 = openmc.Cell(cell_id=12)
c12.region = +s5 & -s6 & +s36 & -s38
c12.fill = upper_rad_ref
root.add_cells((c1, c2, c3, c4, c5, c6, c7, c8, c9, c10, c11, c12))
# Assign root universe to geometry
self.geometry.root_universe = root
def build_default_settings(self):
self.settings.batches = 10
self.settings.inactive = 5
self.settings.particles = 100
self.settings.source = Source(space=Box(
[-160, -160, -183], [160, 160, 183]))
def build_defualt_plots(self):
plot = openmc.Plot()
plot.filename = 'mat'
plot.origin = (125, 125, 0)
plot.width = (250, 250)
plot.pixels = (3000, 3000)
plot.color_by = 'material'
self.plots.add_plot(plot)
class PinCellInputSet(object):
def __init__(self):
self.settings = openmc.Settings()
self.materials = openmc.Materials()
self.geometry = openmc.Geometry()
self.tallies = None
self.plots = None
def export(self):
self.settings.export_to_xml()
self.materials.export_to_xml()
self.geometry.export_to_xml()
if self.tallies is not None:
self.tallies.export_to_xml()
if self.plots is not None:
self.plots.export_to_xml()
def build_default_materials_and_geometry(self):
# Define materials.
fuel = openmc.Material(name='Fuel')
fuel.set_density('g/cm3', 10.29769)
fuel.add_nuclide("U234", 4.4843e-6)
fuel.add_nuclide("U235", 5.5815e-4)
fuel.add_nuclide("U238", 2.2408e-2)
fuel.add_nuclide("O16", 4.5829e-2)
clad = openmc.Material(name='Cladding')
clad.set_density('g/cm3', 6.55)
clad.add_nuclide("Zr90", 2.1827e-2)
clad.add_nuclide("Zr91", 4.7600e-3)
clad.add_nuclide("Zr92", 7.2758e-3)
clad.add_nuclide("Zr94", 7.3734e-3)
clad.add_nuclide("Zr96", 1.1879e-3)
hot_water = openmc.Material(name='Hot borated water')
hot_water.set_density('g/cm3', 0.740582)
hot_water.add_nuclide("H1", 4.9457e-2)
hot_water.add_nuclide("O16", 2.4672e-2)
hot_water.add_nuclide("B10", 8.0042e-6)
hot_water.add_nuclide("B11", 3.2218e-5)
hot_water.add_s_alpha_beta('c_H_in_H2O')
# Define the materials file.
self.materials += (fuel, clad, hot_water)
# Instantiate ZCylinder surfaces
fuel_or = openmc.ZCylinder(x0=0, y0=0, R=0.39218, name='Fuel OR')
clad_or = openmc.ZCylinder(x0=0, y0=0, R=0.45720, name='Clad OR')
left = openmc.XPlane(x0=-0.63, name='left', boundary_type='reflective')
right = openmc.XPlane(x0=0.63, name='right', boundary_type='reflective')
bottom = openmc.YPlane(y0=-0.63, name='bottom',
boundary_type='reflective')
top = openmc.YPlane(y0=0.63, name='top', boundary_type='reflective')
# Instantiate Cells
fuel_pin = openmc.Cell(name='cell 1', fill=fuel)
cladding = openmc.Cell(name='cell 3', fill=clad)
water = openmc.Cell(name='cell 2', fill=hot_water)
# Use surface half-spaces to define regions
fuel_pin.region = -fuel_or
cladding.region = +fuel_or & -clad_or
water.region = +clad_or & +left & -right & +bottom & -top
# Instantiate Universe
root = openmc.Universe(universe_id=0, name='root universe')
# Register Cells with Universe
root.add_cells([fuel_pin, cladding, water])
# Instantiate a Geometry, register the root Universe, and export to XML
self.geometry.root_universe = root
def build_default_settings(self):
self.settings.batches = 10
self.settings.inactive = 5
self.settings.particles = 100
self.settings.source = Source(space=Box([-0.63, -0.63, -1],
[0.63, 0.63, 1],
only_fissionable=True))
def build_defualt_plots(self):
plot = openmc.Plot()
plot.filename = 'mat'
plot.origin = (0.0, 0.0, 0)
plot.width = (1.26, 1.26)
plot.pixels = (300, 300)
plot.color_by = 'material'
self.plots.add_plot(plot)
class AssemblyInputSet(object):
def __init__(self):
self.settings = openmc.Settings()
self.materials = openmc.Materials()
self.geometry = openmc.Geometry()
self.tallies = None
self.plots = None
def export(self):
self.settings.export_to_xml()
self.materials.export_to_xml()
self.geometry.export_to_xml()
if self.tallies is not None:
self.tallies.export_to_xml()
if self.plots is not None:
self.plots.export_to_xml()
def build_default_materials_and_geometry(self):
# Define materials.
fuel = openmc.Material(name='Fuel')
fuel.set_density('g/cm3', 10.29769)
fuel.add_nuclide("U234", 4.4843e-6)
fuel.add_nuclide("U235", 5.5815e-4)
fuel.add_nuclide("U238", 2.2408e-2)
fuel.add_nuclide("O16", 4.5829e-2)
clad = openmc.Material(name='Cladding')
clad.set_density('g/cm3', 6.55)
clad.add_nuclide("Zr90", 2.1827e-2)
clad.add_nuclide("Zr91", 4.7600e-3)
clad.add_nuclide("Zr92", 7.2758e-3)
clad.add_nuclide("Zr94", 7.3734e-3)
clad.add_nuclide("Zr96", 1.1879e-3)
hot_water = openmc.Material(name='Hot borated water')
hot_water.set_density('g/cm3', 0.740582)
hot_water.add_nuclide("H1", 4.9457e-2)
hot_water.add_nuclide("O16", 2.4672e-2)
hot_water.add_nuclide("B10", 8.0042e-6)
hot_water.add_nuclide("B11", 3.2218e-5)
hot_water.add_s_alpha_beta('c_H_in_H2O')
# Define the materials file.
self.materials += (fuel, clad, hot_water)
# Instantiate ZCylinder surfaces
fuel_or = openmc.ZCylinder(x0=0, y0=0, R=0.39218, name='Fuel OR')
clad_or = openmc.ZCylinder(x0=0, y0=0, R=0.45720, name='Clad OR')
# Create boundary planes to surround the geometry
min_x = openmc.XPlane(x0=-10.71, boundary_type='reflective')
max_x = openmc.XPlane(x0=+10.71, boundary_type='reflective')
min_y = openmc.YPlane(y0=-10.71, boundary_type='reflective')
max_y = openmc.YPlane(y0=+10.71, boundary_type='reflective')
# Create a Universe to encapsulate a fuel pin
fuel_pin_universe = openmc.Universe(name='Fuel Pin')
# Create fuel Cell
fuel_cell = openmc.Cell(name='fuel')
fuel_cell.fill = fuel
fuel_cell.region = -fuel_or
fuel_pin_universe.add_cell(fuel_cell)
# Create a clad Cell
clad_cell = openmc.Cell(name='clad')
clad_cell.fill = clad
clad_cell.region = +fuel_or & -clad_or
fuel_pin_universe.add_cell(clad_cell)
# Create a moderator Cell
hot_water_cell = openmc.Cell(name='hot water')
hot_water_cell.fill = hot_water
hot_water_cell.region = +clad_or
fuel_pin_universe.add_cell(hot_water_cell)
# Create a Universe to encapsulate a control rod guide tube
guide_tube_universe = openmc.Universe(name='Guide Tube')
# Create guide tube inner Cell
gt_inner_cell = openmc.Cell(name='guide tube inner water')
gt_inner_cell.fill = hot_water
gt_inner_cell.region = -fuel_or
guide_tube_universe.add_cell(gt_inner_cell)
# Create a clad Cell
gt_clad_cell = openmc.Cell(name='guide tube clad')
gt_clad_cell.fill = clad
gt_clad_cell.region = +fuel_or & -clad_or
guide_tube_universe.add_cell(gt_clad_cell)
# Create a guide tube outer Cell
gt_outer_cell = openmc.Cell(name='guide tube outer water')
gt_outer_cell.fill = hot_water
gt_outer_cell.region = +clad_or
guide_tube_universe.add_cell(gt_outer_cell)
# Create fuel assembly Lattice
assembly = openmc.RectLattice(name='Fuel Assembly')
assembly.pitch = (1.26, 1.26)
assembly.lower_left = [-1.26 * 17. / 2.0] * 2
# Create array indices for guide tube locations in lattice
template_x = np.array([5, 8, 11, 3, 13, 2, 5, 8, 11, 14, 2, 5, 8,
11, 14, 2, 5, 8, 11, 14, 3, 13, 5, 8, 11])
template_y = np.array([2, 2, 2, 3, 3, 5, 5, 5, 5, 5, 8, 8, 8, 8,
8, 11, 11, 11, 11, 11, 13, 13, 14, 14, 14])
# Initialize an empty 17x17 array of the lattice universes
universes = np.empty((17, 17), dtype=openmc.Universe)
# Fill the array with the fuel pin and guide tube universes
universes[:,:] = fuel_pin_universe
universes[template_x, template_y] = guide_tube_universe
# Store the array of universes in the lattice
assembly.universes = universes
# Create root Cell
root_cell = openmc.Cell(name='root cell')
root_cell.fill = assembly
# Add boundary planes
root_cell.region = +min_x & -max_x & +min_y & -max_y
# Create root Universe
root_universe = openmc.Universe(universe_id=0, name='root universe')
root_universe.add_cell(root_cell)
# Instantiate a Geometry, register the root Universe, and export to XML
self.geometry.root_universe = root_universe
def build_default_settings(self):
self.settings.batches = 10
self.settings.inactive = 5
self.settings.particles = 100
self.settings.source = Source(space=Box([-10.71, -10.71, -1],
[10.71, 10.71, 1],
only_fissionable=True))
def build_defualt_plots(self):
plot = openmc.Plot()
plot.filename = 'mat'
plot.origin = (0.0, 0.0, 0)
plot.width = (21.42, 21.42)
plot.pixels = (300, 300)
plot.color_by = 'material'
self.plots.add_plot(plot)
class MGInputSet(InputSet):
def build_default_materials_and_geometry(self, reps=None, as_macro=True):
# Define materials needed for 1D/1G slab problem
mat_names = ['uo2', 'clad', 'lwtr']
mgxs_reps = ['ang', 'ang_mu', 'iso', 'iso_mu']
if reps is None:
reps = mgxs_reps
xs = []
mats = []
i = 0
for mat in mat_names:
for rep in reps:
i += 1
if as_macro:
xs.append(openmc.Macroscopic(mat + '_' + rep))
mats.append(openmc.Material(name=str(i)))
mats[-1].set_density('macro', 1.)
mats[-1].add_macroscopic(xs[-1])
else:
xs.append(openmc.Nuclide(mat + '_' + rep))
mats.append(openmc.Material(name=str(i)))
mats[-1].set_density('atom/b-cm', 1.)
mats[-1].add_nuclide(xs[-1].name, 1.0, 'ao')
# Define the materials file
self.xs_data = xs
self.materials += mats
self.materials.cross_sections = "../1d_mgxs.h5"
# Define surfaces.
# Assembly/Problem Boundary
left = openmc.XPlane(x0=0.0, boundary_type='reflective')
right = openmc.XPlane(x0=10.0, boundary_type='reflective')
bottom = openmc.YPlane(y0=0.0, boundary_type='reflective')
top = openmc.YPlane(y0=10.0, boundary_type='reflective')
# for each material add a plane
planes = [openmc.ZPlane(z0=0.0, boundary_type='reflective')]
dz = round(5. / float(len(mats)), 4)
for i in range(len(mats) - 1):
planes.append(openmc.ZPlane(z0=dz * float(i + 1)))
planes.append(openmc.ZPlane(z0=5.0, boundary_type='reflective'))
# Define cells for each material
cells = []
xy = +left & -right & +bottom & -top
for i, mat in enumerate(mats):
cells.append(openmc.Cell())
cells[-1].region = xy & +planes[i] & -planes[i + 1]
cells[-1].fill = mat
# Define root universe.
root = openmc.Universe(universe_id=0, name='root universe')
root.add_cells(cells)
# Assign root universe to geometry
self.geometry.root_universe = root
def build_default_settings(self):
self.settings.batches = 10
self.settings.inactive = 5
self.settings.particles = 100
self.settings.source = Source(space=Box([0.0, 0.0, 0.0],
[10.0, 10.0, 5.]))
self.settings.energy_mode = "multi-group"
def build_defualt_plots(self):
plot = openmc.Plot()
plot.filename = 'mat'
plot.origin = (5.0, 5.0, 2.5)
plot.width = (2.5, 2.5)
plot.basis = 'xz'
plot.pixels = (3000, 3000)
plot.color_by = 'material'
self.plots.add_plot(plot)

4
tests/run_tests.py
View file

@ -184,8 +184,8 @@ class Test(object):
build_str += "-Ddebug=ON "
if self.optimize:
build_str += "-Doptimize=ON "
if self.openmp:
build_str += "-Dopenmp=ON "
if not self.openmp:
build_str += "-Dopenmp=OFF "
if self.coverage:
build_str += "-Dcoverage=ON "
self.build_opts = build_str

6
tests/test_asymmetric_lattice/inputs_true.dat
View file

@ -56,7 +56,7 @@
</geometry>
<?xml version='1.0' encoding='utf-8'?>
<materials>
<material id="1" name="Fuel">
<material id="1" name="UOX fuel">
<density units="g/cm3" value="10.062" />
<nuclide ao="4.9476e-06" name="U234" />
<nuclide ao="0.00048218" name="U235" />
@ -105,7 +105,7 @@
<nuclide name="Zr96" wo="0.0216912612" />
<sab name="c_H_in_H2O" />
</material>
<material id="2" name="Cladding">
<material id="2" name="Zircaloy">
<density units="g/cm3" value="5.77" />
<nuclide ao="0.5145" name="Zr90" />
<nuclide ao="0.1122" name="Zr91" />
@ -157,7 +157,7 @@
<nuclide name="Cr52" wo="0.145407678031" />
<sab name="c_H_in_H2O" />
</material>
<material id="7" name="Upper radial reflector /Top plate region">
<material id="7" name="Upper radial reflector / Top plate region">
<density units="g/cm3" value="4.28" />
<nuclide name="H1" wo="0.0086117" />
<nuclide name="O16" wo="0.0683369" />

30
tests/test_asymmetric_lattice/test_asymmetric_lattice.py
View file

@ -10,15 +10,11 @@ import openmc
class AsymmetricLatticeTestHarness(PyAPITestHarness):
def _build_inputs(self):
"""Build an axis-asymmetric lattice of fuel assemblies"""
# Build full core geometry from underlying input set
self._input_set.build_default_materials_and_geometry()
def __init__(self, *args, **kwargs):
super(AsymmetricLatticeTestHarness, self).__init__(*args, **kwargs)
# Extract universes encapsulating fuel and water assemblies
geometry = self._input_set.geometry
geometry = self._model.geometry
water = geometry.get_universes_by_name('water assembly (hot)')[0]
fuel = geometry.get_universes_by_name('fuel assembly (hot)')[0]
@ -46,7 +42,7 @@ class AsymmetricLatticeTestHarness(PyAPITestHarness):
root_univ.add_cell(root_cell)
# Over-ride geometry in the input set with this 3x3 lattice
self._input_set.geometry.root_universe = root_univ
self._model.geometry.root_universe = root_univ
# Initialize a "distribcell" filter for the fuel pin cell
distrib_filter = openmc.DistribcellFilter(27)
@ -56,22 +52,12 @@ class AsymmetricLatticeTestHarness(PyAPITestHarness):
tally.filters.append(distrib_filter)
tally.scores.append('nu-fission')
# Initialize the tallies file
tallies_file = openmc.Tallies([tally])
# Assign the tallies file to the input set
self._input_set.tallies = tallies_file
# Build default settings
self._input_set.build_default_settings()
self._model.tallies.append(tally)
# Specify summary output and correct source sampling box
source = openmc.Source(space=openmc.stats.Box([-32, -32, 0], [32, 32, 32]))
source.space.only_fissionable = True
self._input_set.settings.source = source
# Write input XML files
self._input_set.export()
self._model.settings.source = openmc.Source(space=openmc.stats.Box(
[-32, -32, 0], [32, 32, 32], only_fissionable = True))
def _get_results(self, hash_output=True):
"""Digest info in statepoint and summary and return as a string."""
@ -105,5 +91,5 @@ class AsymmetricLatticeTestHarness(PyAPITestHarness):
if __name__ == '__main__':
harness = AsymmetricLatticeTestHarness('statepoint.10.h5', True)
harness = AsymmetricLatticeTestHarness('statepoint.10.h5')
harness.main()

2
tests/test_cmfd_feed/test_cmfd_feed.py
View file

@ -7,5 +7,5 @@ from testing_harness import CMFDTestHarness
if __name__ == '__main__':
harness = CMFDTestHarness('statepoint.20.*', True)
harness = CMFDTestHarness('statepoint.20.h5')
harness.main()

2
tests/test_cmfd_nofeed/test_cmfd_nofeed.py
View file

@ -7,5 +7,5 @@ from testing_harness import CMFDTestHarness
if __name__ == '__main__':
harness = CMFDTestHarness('statepoint.20.*', True)
harness = CMFDTestHarness('statepoint.20.h5')
harness.main()

2
tests/test_complex_cell/test_complex_cell.py
View file

@ -6,5 +6,5 @@ from testing_harness import TestHarness
if __name__ == '__main__':
harness = TestHarness('statepoint.10.*', True)
harness = TestHarness('statepoint.10.h5')
harness.main()

2
tests/test_confidence_intervals/test_confidence_intervals.py
View file

@ -7,5 +7,5 @@ from testing_harness import TestHarness
if __name__ == '__main__':
harness = TestHarness('statepoint.10.*', True)
harness = TestHarness('statepoint.10.h5')
harness.main()

2
tests/test_create_fission_neutrons/test_create_fission_neutrons.py
View file

@ -70,5 +70,5 @@ class CreateFissionNeutronsTestHarness(PyAPITestHarness):
if __name__ == '__main__':
harness = CreateFissionNeutronsTestHarness('statepoint.10.h5', True)
harness = CreateFissionNeutronsTestHarness('statepoint.10.h5')
harness.main()

2
tests/test_density/test_density.py
View file

@ -7,5 +7,5 @@ from testing_harness import TestHarness
if __name__ == '__main__':
harness = TestHarness('statepoint.10.*')
harness = TestHarness('statepoint.10.h5')
harness.main()

6
tests/test_diff_tally/inputs_true.dat
View file

@ -148,7 +148,7 @@
</geometry>
<?xml version='1.0' encoding='utf-8'?>
<materials>
<material id="1" name="Fuel">
<material id="1" name="UOX fuel">
<density units="g/cm3" value="10.062" />
<nuclide ao="4.9476e-06" name="U234" />
<nuclide ao="0.00048218" name="U235" />
@ -197,7 +197,7 @@
<nuclide name="Zr96" wo="0.0216912612" />
<sab name="c_H_in_H2O" />
</material>
<material id="2" name="Cladding">
<material id="2" name="Zircaloy">
<density units="g/cm3" value="5.77" />
<nuclide ao="0.5145" name="Zr90" />
<nuclide ao="0.1122" name="Zr91" />
@ -249,7 +249,7 @@
<nuclide name="Cr52" wo="0.145407678031" />
<sab name="c_H_in_H2O" />
</material>
<material id="7" name="Upper radial reflector /Top plate region">
<material id="7" name="Upper radial reflector / Top plate region">
<density units="g/cm3" value="4.28" />
<nuclide name="H1" wo="0.0086117" />
<nuclide name="O16" wo="0.0683369" />

31
tests/test_diff_tally/test_diff_tally.py
View file

@ -11,19 +11,16 @@ from testing_harness import PyAPITestHarness
import openmc
class DiffTallyTestHarness(PyAPITestHarness):
def _build_inputs(self):
# Build default materials/geometry
self._input_set.build_default_materials_and_geometry()
def __init__(self, *args, **kwargs):
super(DiffTallyTestHarness, self).__init__(*args, **kwargs)
# Set settings explicitly
self._input_set.settings.batches = 3
self._input_set.settings.inactive = 0
self._input_set.settings.particles = 100
self._input_set.settings.source = openmc.Source(space=openmc.stats.Box(
self._model.settings.batches = 3
self._model.settings.inactive = 0
self._model.settings.particles = 100
self._model.settings.source = openmc.Source(space=openmc.stats.Box(
[-160, -160, -183], [160, 160, 183]))
self._input_set.settings.temperature['multipole'] = True
self._input_set.tallies = openmc.Tallies()
self._model.settings.temperature['multipole'] = True
filt_mats = openmc.MaterialFilter((1, 3))
filt_eout = openmc.EnergyoutFilter((0.0, 0.625, 20.0e6))
@ -64,7 +61,7 @@ class DiffTallyTestHarness(PyAPITestHarness):
t.add_score('flux')
t.add_filter(filt_mats)
t.derivative = derivs[i]
self._input_set.tallies.append(t)
self._model.tallies.append(t)
# Cover supported scores with a collision estimator.
for i in range(5):
@ -78,7 +75,7 @@ class DiffTallyTestHarness(PyAPITestHarness):
t.add_nuclide('total')
t.add_nuclide('U235')
t.derivative = derivs[i]
self._input_set.tallies.append(t)
self._model.tallies.append(t)
# Cover an analog estimator.
for i in range(5):
@ -87,7 +84,7 @@ class DiffTallyTestHarness(PyAPITestHarness):
t.add_filter(filt_mats)
t.estimator = 'analog'
t.derivative = derivs[i]
self._input_set.tallies.append(t)
self._model.tallies.append(t)
# Energyout filter and total nuclide for the density derivatives.
for i in range(2):
@ -99,7 +96,7 @@ class DiffTallyTestHarness(PyAPITestHarness):
t.add_nuclide('total')
t.add_nuclide('U235')
t.derivative = derivs[i]
self._input_set.tallies.append(t)
self._model.tallies.append(t)
# Energyout filter without total nuclide for other derivatives.
for i in range(2, 5):
@ -110,9 +107,7 @@ class DiffTallyTestHarness(PyAPITestHarness):
t.add_filter(filt_eout)
t.add_nuclide('U235')
t.derivative = derivs[i]
self._input_set.tallies.append(t)
self._input_set.export()
self._model.tallies.append(t)
def _get_results(self):
# Read the statepoint and summary files.
@ -131,5 +126,5 @@ class DiffTallyTestHarness(PyAPITestHarness):
if __name__ == '__main__':
harness = DiffTallyTestHarness('statepoint.3.h5', True)
harness = DiffTallyTestHarness('statepoint.3.h5')
harness.main()

3
tests/test_distribmat/inputs_true.dat
View file

@ -46,9 +46,6 @@
<parameters>-1 -1 -1 1 1 1</parameters>
</space>
</source>
<output>
<summary>true</summary>
</output>
</settings>
<?xml version='1.0' encoding='utf-8'?>
<plots>

75
tests/test_distribmat/test_distribmat.py
View file

@ -5,8 +5,6 @@ import sys
sys.path.insert(0, os.pardir)
from testing_harness import TestHarness, PyAPITestHarness
import openmc
from openmc.stats import Box
from openmc.source import Source
class DistribmatTestHarness(PyAPITestHarness):
@ -31,25 +29,18 @@ class DistribmatTestHarness(PyAPITestHarness):
mats_file = openmc.Materials([moderator, dense_fuel, light_fuel])
mats_file.export_to_xml()
####################
# Geometry
####################
c1 = openmc.Cell(cell_id=1)
c1.fill = moderator
mod_univ = openmc.Universe(universe_id=1)
mod_univ.add_cell(c1)
c1 = openmc.Cell(cell_id=1, fill=moderator)
mod_univ = openmc.Universe(universe_id=1, cells=[c1])
r0 = openmc.ZCylinder(R=0.3)
c11 = openmc.Cell(cell_id=11)
c11.region = -r0
c11 = openmc.Cell(cell_id=11, region=-r0)
c11.fill = [dense_fuel, light_fuel, None, dense_fuel]
c12 = openmc.Cell(cell_id=12)
c12.region = +r0
c12.fill = moderator
fuel_univ = openmc.Universe(universe_id=11)
fuel_univ.add_cells((c11, c12))
c12 = openmc.Cell(cell_id=12, region=+r0, fill=moderator)
fuel_univ = openmc.Universe(universe_id=11, cells=[c11, c12])
lat = openmc.RectLattice(lattice_id=101)
lat.lower_left = [-2.0, -2.0]
@ -63,17 +54,13 @@ class DistribmatTestHarness(PyAPITestHarness):
y1 = openmc.YPlane(y0=3.0)
for s in [x0, x1, y0, y1]:
s.boundary_type = 'reflective'
c101 = openmc.Cell(cell_id=101)
c101 = openmc.Cell(cell_id=101, fill=lat)
c101.region = +x0 & -x1 & +y0 & -y1
c101.fill = lat
root_univ = openmc.Universe(universe_id=0)
root_univ.add_cell(c101)
root_univ = openmc.Universe(universe_id=0, cells=[c101])
geometry = openmc.Geometry()
geometry.root_universe = root_univ
geometry = openmc.Geometry(root_univ)
geometry.export_to_xml()
####################
# Settings
####################
@ -82,36 +69,32 @@ class DistribmatTestHarness(PyAPITestHarness):
sets_file.batches = 5
sets_file.inactive = 0
sets_file.particles = 1000
sets_file.source = Source(space=Box([-1, -1, -1], [1, 1, 1]))
sets_file.output = {'summary': True}
sets_file.source = openmc.Source(space=openmc.stats.Box(
[-1, -1, -1], [1, 1, 1]))
sets_file.export_to_xml()
####################
# Plots
####################
plots_file = openmc.Plots()
plot1 = openmc.Plot(plot_id=1)
plot1.basis = 'xy'
plot1.color_by = 'cell'
plot1.filename = 'cellplot'
plot1.origin = (0, 0, 0)
plot1.width = (7, 7)
plot1.pixels = (400, 400)
plot = openmc.Plot(plot_id=1)
plot.basis = 'xy'
plot.color_by = 'cell'
plot.filename = 'cellplot'
plot.origin = (0, 0, 0)
plot.width = (7, 7)
plot.pixels = (400, 400)
plots_file.add_plot(plot)
plot2 = openmc.Plot(plot_id=2)
plot2.basis = 'xy'
plot2.color_by = 'material'
plot2.filename = 'matplot'
plot2.origin = (0, 0, 0)
plot2.width = (7, 7)
plot2.pixels = (400, 400)
plot = openmc.Plot(plot_id=2)
plot.basis = 'xy'
plot.color_by = 'material'
plot.filename = 'matplot'
plot.origin = (0, 0, 0)
plot.width = (7, 7)
plot.pixels = (400, 400)
plots_file.add_plot(plot)
plots_file.export_to_xml()
plots = openmc.Plots([plot1, plot2])
plots.export_to_xml()
def _get_results(self):
outstr = super(DistribmatTestHarness, self)._get_results()
@ -119,12 +102,6 @@ class DistribmatTestHarness(PyAPITestHarness):
outstr += str(su.geometry.get_all_cells()[11])
return outstr
def _cleanup(self):
f = os.path.join(os.getcwd(), 'plots.xml')
if os.path.exists(f):
os.remove(f)
super(DistribmatTestHarness, self)._cleanup()
if __name__ == '__main__':
harness = DistribmatTestHarness('statepoint.5.h5')

2
tests/test_eigenvalue_genperbatch/test_eigenvalue_genperbatch.py
View file

@ -7,5 +7,5 @@ from testing_harness import TestHarness
if __name__ == '__main__':
harness = TestHarness('statepoint.7.*')
harness = TestHarness('statepoint.7.h5')
harness.main()

2
tests/test_eigenvalue_no_inactive/test_eigenvalue_no_inactive.py
View file

@ -7,5 +7,5 @@ from testing_harness import TestHarness
if __name__ == '__main__':
harness = TestHarness('statepoint.10.*')
harness = TestHarness('statepoint.10.h5')
harness.main()

2
tests/test_energy_cutoff/test_energy_cutoff.py
View file

@ -74,5 +74,5 @@ class EnergyCutoffTestHarness(PyAPITestHarness):
if __name__ == '__main__':
harness = EnergyCutoffTestHarness('statepoint.10.h5', True)
harness = EnergyCutoffTestHarness('statepoint.10.h5')
harness.main()

2
tests/test_energy_grid/test_energy_grid.py
View file

@ -7,5 +7,5 @@ from testing_harness import TestHarness
if __name__ == '__main__':
harness = TestHarness('statepoint.10.*')
harness = TestHarness('statepoint.10.h5')
harness.main()

2
tests/test_energy_laws/test_energy_laws.py
View file

@ -26,5 +26,5 @@ from testing_harness import TestHarness
if __name__ == '__main__':
harness = TestHarness('statepoint.10.*')
harness = TestHarness('statepoint.10.h5')
harness.main()

22
tests/test_entropy/test_entropy.py
View file

@ -5,7 +5,7 @@ import os
import sys
sys.path.insert(0, os.pardir)
from testing_harness import TestHarness
from openmc.statepoint import StatePoint
from openmc import StatePoint
class EntropyTestHarness(TestHarness):
@ -13,21 +13,19 @@ class EntropyTestHarness(TestHarness):
"""Digest info in the statepoint and return as a string."""
# Read the statepoint file.
statepoint = glob.glob(os.path.join(os.getcwd(), self._sp_name))[0]
sp = StatePoint(statepoint)
with StatePoint(statepoint) as sp:
# Write out k-combined.
outstr = 'k-combined:\n'
outstr += '{0:12.6E} {1:12.6E}\n'.format(*sp.k_combined)
# Write out k-combined.
outstr = 'k-combined:\n'
form = '{0:12.6E} {1:12.6E}\n'
outstr += form.format(sp.k_combined[0], sp.k_combined[1])
# Write out entropy data.
outstr += 'entropy:\n'
results = ['{0:12.6E}'.format(x) for x in sp.entropy]
outstr += '\n'.join(results) + '\n'
# Write out entropy data.
outstr += 'entropy:\n'
results = ['{0:12.6E}'.format(x) for x in sp.entropy]
outstr += '\n'.join(results) + '\n'
return outstr
if __name__ == '__main__':
harness = EntropyTestHarness('statepoint.10.*')
harness = EntropyTestHarness('statepoint.10.h5')
harness.main()

6
tests/test_filter_distribcell/test_filter_distribcell.py
View file

@ -10,7 +10,7 @@ from testing_harness import *
class DistribcellTestHarness(TestHarness):
def __init__(self):
super(DistribcellTestHarness, self).__init__(None, True)
super(DistribcellTestHarness, self).__init__(None)
def execute_test(self):
"""Run OpenMC with the appropriate arguments and check the outputs."""
@ -41,8 +41,8 @@ class DistribcellTestHarness(TestHarness):
base_dir = os.getcwd()
try:
dirs = ('case-1', '../case-2', '../case-3', '../case-4')
sps = ('statepoint.1.*', 'statepoint.1.*', 'statepoint.3.*',
'statepoint.1.*')
sps = ('statepoint.1.h5', 'statepoint.1.h5', 'statepoint.3.h5',
'statepoint.1.h5')
tallies_out_present = (True, True, False, True)
hash_out = (False, False, True, False)
for i in range(len(dirs)):

6
tests/test_filter_energyfun/inputs_true.dat
View file

@ -148,7 +148,7 @@
</geometry>
<?xml version='1.0' encoding='utf-8'?>
<materials>
<material id="1" name="Fuel">
<material id="1" name="UOX fuel">
<density units="g/cm3" value="10.062" />
<nuclide ao="4.9476e-06" name="U234" />
<nuclide ao="0.00048218" name="U235" />
@ -197,7 +197,7 @@
<nuclide name="Zr96" wo="0.0216912612" />
<sab name="c_H_in_H2O" />
</material>
<material id="2" name="Cladding">
<material id="2" name="Zircaloy">
<density units="g/cm3" value="5.77" />
<nuclide ao="0.5145" name="Zr90" />
<nuclide ao="0.1122" name="Zr91" />
@ -250,7 +250,7 @@
<nuclide name="Cr52" wo="0.145407678031" />
<sab name="c_H_in_H2O" />
</material>
<material id="7" name="Upper radial reflector /Top plate region">
<material id="7" name="Upper radial reflector / Top plate region">
<density units="g/cm3" value="4.28" />
<nuclide name="H1" wo="0.0086117" />
<nuclide name="O16" wo="0.0683369" />

16
tests/test_filter_energyfun/test_filter_energyfun.py
View file

@ -3,20 +3,17 @@
import os
import sys
import glob
import hashlib
sys.path.insert(0, os.pardir)
from testing_harness import PyAPITestHarness
import openmc
class FilterEnergyFunHarness(PyAPITestHarness):
def _build_inputs(self):
# Build the default material, geometry, settings inputs.
self._input_set.build_default_materials_and_geometry()
self._input_set.build_default_settings()
def __init__(self, *args, **kwargs):
super(FilterEnergyFunHarness, self).__init__(*args, **kwargs)
# Add Am241 to the fuel.
self._input_set.materials[1].add_nuclide('Am241', 1e-7)
self._model.materials[1].add_nuclide('Am241', 1e-7)
# Define Am242m / Am242 branching ratio from ENDF/B-VII.1 data.
x = [1e-5, 3.69e-1, 1e3, 1e5, 6e5, 1e6, 2e6, 4e6, 3e7]
@ -37,10 +34,7 @@ class FilterEnergyFunHarness(PyAPITestHarness):
t.scores = ['(n,gamma)']
t.nuclides = ['Am241']
tallies[1].filters = [filt1]
self._input_set.tallies = openmc.Tallies(tallies)
# Export inputs to xml.
self._input_set.export()
self._model.tallies = tallies
def _get_results(self):
# Read the statepoint file.
@ -55,5 +49,5 @@ class FilterEnergyFunHarness(PyAPITestHarness):
if __name__ == '__main__':
harness = FilterEnergyFunHarness('statepoint.10.h5', True)
harness = FilterEnergyFunHarness('statepoint.10.h5')
harness.main()

9
tests/test_filter_mesh/inputs_true.dat
View file

@ -148,7 +148,7 @@
</geometry>
<?xml version='1.0' encoding='utf-8'?>
<materials>
<material id="1" name="Fuel">
<material id="1" name="UOX fuel">
<density units="g/cm3" value="10.062" />
<nuclide ao="4.9476e-06" name="U234" />
<nuclide ao="0.00048218" name="U235" />
@ -197,7 +197,7 @@
<nuclide name="Zr96" wo="0.0216912612" />
<sab name="c_H_in_H2O" />
</material>
<material id="2" name="Cladding">
<material id="2" name="Zircaloy">
<density units="g/cm3" value="5.77" />
<nuclide ao="0.5145" name="Zr90" />
<nuclide ao="0.1122" name="Zr91" />
@ -249,7 +249,7 @@
<nuclide name="Cr52" wo="0.145407678031" />
<sab name="c_H_in_H2O" />
</material>
<material id="7" name="Upper radial reflector /Top plate region">
<material id="7" name="Upper radial reflector / Top plate region">
<density units="g/cm3" value="4.28" />
<nuclide name="H1" wo="0.0086117" />
<nuclide name="O16" wo="0.0683369" />
@ -306,9 +306,6 @@
<parameters>-160 -160 -183 160 160 183</parameters>
</space>
</source>
<output>
<summary>true</summary>
</output>
</settings>
<?xml version='1.0' encoding='utf-8'?>
<tallies>

35
tests/test_filter_mesh/test_filter_mesh.py
View file

@ -2,21 +2,14 @@
import os
import sys
import glob
import hashlib
sys.path.insert(0, os.pardir)
from testing_harness import HashedPyAPITestHarness
import openmc
class FilterMeshTestHarness(HashedPyAPITestHarness):
def _build_inputs(self):
# The summary.h5 file needs to be created to read in the tallies
self._input_set.settings.output = {'summary': True}
# Initialize the tallies file
tallies_file = openmc.Tallies()
def __init__(self, *args, **kwargs):
super(FilterMeshTestHarness, self).__init__(*args, **kwargs)
# Initialize Meshes
mesh_1d = openmc.Mesh(mesh_id=1)
@ -38,46 +31,42 @@ class FilterMeshTestHarness(HashedPyAPITestHarness):
mesh_3d.upper_right = [182.07, 182.07, 183.00]
# Initialize the filters
mesh_1d_filter = openmc.MeshFilter(mesh_1d)
mesh_2d_filter = openmc.MeshFilter(mesh_2d)
mesh_3d_filter = openmc.MeshFilter(mesh_3d)
mesh_1d_filter = openmc.MeshFilter(mesh_1d)
mesh_2d_filter = openmc.MeshFilter(mesh_2d)
mesh_3d_filter = openmc.MeshFilter(mesh_3d)
# Initialized the tallies
tally = openmc.Tally(name='tally 1')
tally.filters = [mesh_1d_filter]
tally.scores = ['total']
tallies_file.append(tally)
self._model.tallies.append(tally)
tally = openmc.Tally(name='tally 2')
tally.filters = [mesh_1d_filter]
tally.scores = ['current']
tallies_file.append(tally)
self._model.tallies.append(tally)
tally = openmc.Tally(name='tally 3')
tally.filters = [mesh_2d_filter]
tally.scores = ['total']
tallies_file.append(tally)
self._model.tallies.append(tally)
tally = openmc.Tally(name='tally 4')
tally.filters = [mesh_2d_filter]
tally.scores = ['current']
tallies_file.append(tally)
self._model.tallies.append(tally)
tally = openmc.Tally(name='tally 5')
tally.filters = [mesh_3d_filter]
tally.scores = ['total']
tallies_file.append(tally)
self._model.tallies.append(tally)
tally = openmc.Tally(name='tally 6')
tally.filters = [mesh_3d_filter]
tally.scores = ['current']
tallies_file.append(tally)
# Export tallies to file
self._input_set.tallies = tallies_file
super(FilterMeshTestHarness, self)._build_inputs()
self._model.tallies.append(tally)
if __name__ == '__main__':
harness = FilterMeshTestHarness('statepoint.10.h5', True)
harness = FilterMeshTestHarness('statepoint.10.h5')
harness.main()

24
tests/test_fixed_source/test_fixed_source.py
View file

@ -6,7 +6,7 @@ import sys
import numpy as np
sys.path.insert(0, os.pardir)
from testing_harness import TestHarness
from openmc.statepoint import StatePoint
from openmc import StatePoint
class FixedSourceTestHarness(TestHarness):
@ -14,31 +14,27 @@ class FixedSourceTestHarness(TestHarness):
"""Digest info in the statepoint and return as a string."""
# Read the statepoint file.
statepoint = glob.glob(os.path.join(os.getcwd(), self._sp_name))[0]
sp = StatePoint(statepoint)
# Write out tally data.
outstr = ''
if self._tallies:
tally_num = 1
for tally_ind in sp.tallies:
with StatePoint(statepoint) as sp:
# Write out tally data.
for i, tally_ind in enumerate(sp.tallies):
tally = sp.tallies[tally_ind]
results = np.zeros((tally.sum.size*2, ))
results[0::2] = tally.sum.ravel()
results[1::2] = tally.sum_sq.ravel()
results = ['{0:12.6E}'.format(x) for x in results]
outstr += 'tally ' + str(tally_num) + ':\n'
outstr += 'tally ' + str(i + 1) + ':\n'
outstr += '\n'.join(results) + '\n'
tally_num += 1
gt = sp.global_tallies
outstr += 'leakage:\n'
outstr += '{0:12.6E}'.format(gt[gt['name'] == b'leakage'][0]['sum']) + '\n'
outstr += '{0:12.6E}'.format(gt[gt['name'] == b'leakage'][0]['sum_sq']) + '\n'
gt = sp.global_tallies
outstr += 'leakage:\n'
outstr += '{0:12.6E}'.format(gt[gt['name'] == b'leakage'][0]['sum']) + '\n'
outstr += '{0:12.6E}'.format(gt[gt['name'] == b'leakage'][0]['sum_sq']) + '\n'
return outstr
if __name__ == '__main__':
harness = FixedSourceTestHarness('statepoint.10.*', True)
harness = FixedSourceTestHarness('statepoint.10.h5')
harness.main()

2
tests/test_infinite_cell/test_infinite_cell.py
View file

@ -7,5 +7,5 @@ from testing_harness import TestHarness
if __name__ == '__main__':
harness = TestHarness('statepoint.10.*')
harness = TestHarness('statepoint.10.h5')
harness.main()

6
tests/test_iso_in_lab/inputs_true.dat
View file

@ -148,7 +148,7 @@
</geometry>
<?xml version='1.0' encoding='utf-8'?>
<materials>
<material id="1" name="Fuel">
<material id="1" name="UOX fuel">
<density units="g/cm3" value="10.062" />
<nuclide ao="4.9476e-06" name="U234" scattering="iso-in-lab" />
<nuclide ao="0.00048218" name="U235" scattering="iso-in-lab" />
@ -197,7 +197,7 @@
<nuclide name="Zr96" scattering="iso-in-lab" wo="0.0216912612" />
<sab name="c_H_in_H2O" />
</material>
<material id="2" name="Cladding">
<material id="2" name="Zircaloy">
<density units="g/cm3" value="5.77" />
<nuclide ao="0.5145" name="Zr90" scattering="iso-in-lab" />
<nuclide ao="0.1122" name="Zr91" scattering="iso-in-lab" />
@ -249,7 +249,7 @@
<nuclide name="Cr52" scattering="iso-in-lab" wo="0.145407678031" />
<sab name="c_H_in_H2O" />
</material>
<material id="7" name="Upper radial reflector /Top plate region">
<material id="7" name="Upper radial reflector / Top plate region">
<density units="g/cm3" value="4.28" />
<nuclide name="H1" scattering="iso-in-lab" wo="0.0086117" />
<nuclide name="O16" scattering="iso-in-lab" wo="0.0683369" />

19
tests/test_iso_in_lab/test_iso_in_lab.py
View file

@ -2,25 +2,12 @@
import os
import sys
import glob
import hashlib
sys.path.insert(0, os.pardir)
from testing_harness import PyAPITestHarness
import openmc
import openmc.mgxs
class IsoInLabTestHarness(PyAPITestHarness):
def _build_inputs(self):
"""Write input XML files with iso-in-lab scattering."""
self._input_set.build_default_materials_and_geometry()
self._input_set.build_default_settings()
self._input_set.materials.make_isotropic_in_lab()
self._input_set.export()
if __name__ == '__main__':
harness = IsoInLabTestHarness('statepoint.10.*')
# Force iso-in-lab scattering.
harness = PyAPITestHarness('statepoint.10.h5')
harness._model.materials.make_isotropic_in_lab()
harness.main()

2
tests/test_lattice/test_lattice.py
View file

@ -7,5 +7,5 @@ from testing_harness import TestHarness
if __name__ == '__main__':
harness = TestHarness('statepoint.10.*')
harness = TestHarness('statepoint.10.h5')
harness.main()

2
tests/test_lattice_hex/test_lattice_hex.py
View file

@ -7,5 +7,5 @@ from testing_harness import TestHarness
if __name__ == '__main__':
harness = TestHarness('statepoint.10.*')
harness = TestHarness('statepoint.10.h5')
harness.main()

2
tests/test_lattice_mixed/test_lattice_mixed.py
View file

@ -7,5 +7,5 @@ from testing_harness import TestHarness
if __name__ == '__main__':
harness = TestHarness('statepoint.10.*')
harness = TestHarness('statepoint.10.h5')
harness.main()

2
tests/test_lattice_multiple/test_lattice_multiple.py
View file

@ -7,5 +7,5 @@ from testing_harness import TestHarness
if __name__ == '__main__':
harness = TestHarness('statepoint.10.*')
harness = TestHarness('statepoint.10.h5')
harness.main()

24
tests/test_mg_basic/inputs_true.dat
View file

@ -1,17 +1,17 @@
<?xml version='1.0' encoding='utf-8'?>
<geometry>
<cell id="10000" material="10000" region="10000 -10001 10002 -10003 10004 -10005" universe="0" />
<cell id="10001" material="10001" region="10000 -10001 10002 -10003 10005 -10006" universe="0" />
<cell id="10002" material="10002" region="10000 -10001 10002 -10003 10006 -10007" universe="0" />
<cell id="10003" material="10003" region="10000 -10001 10002 -10003 10007 -10008" universe="0" />
<cell id="10004" material="10004" region="10000 -10001 10002 -10003 10008 -10009" universe="0" />
<cell id="10005" material="10005" region="10000 -10001 10002 -10003 10009 -10010" universe="0" />
<cell id="10006" material="10006" region="10000 -10001 10002 -10003 10010 -10011" universe="0" />
<cell id="10007" material="10007" region="10000 -10001 10002 -10003 10011 -10012" universe="0" />
<cell id="10008" material="10008" region="10000 -10001 10002 -10003 10012 -10013" universe="0" />
<cell id="10009" material="10009" region="10000 -10001 10002 -10003 10013 -10014" universe="0" />
<cell id="10010" material="10010" region="10000 -10001 10002 -10003 10014 -10015" universe="0" />
<cell id="10011" material="10011" region="10000 -10001 10002 -10003 10015 -10016" universe="0" />
<cell id="10000" material="10000" region="10000 -10001 10002 -10003 10004 -10005" universe="10000" />
<cell id="10001" material="10001" region="10000 -10001 10002 -10003 10005 -10006" universe="10000" />
<cell id="10002" material="10002" region="10000 -10001 10002 -10003 10006 -10007" universe="10000" />
<cell id="10003" material="10003" region="10000 -10001 10002 -10003 10007 -10008" universe="10000" />
<cell id="10004" material="10004" region="10000 -10001 10002 -10003 10008 -10009" universe="10000" />
<cell id="10005" material="10005" region="10000 -10001 10002 -10003 10009 -10010" universe="10000" />
<cell id="10006" material="10006" region="10000 -10001 10002 -10003 10010 -10011" universe="10000" />
<cell id="10007" material="10007" region="10000 -10001 10002 -10003 10011 -10012" universe="10000" />
<cell id="10008" material="10008" region="10000 -10001 10002 -10003 10012 -10013" universe="10000" />
<cell id="10009" material="10009" region="10000 -10001 10002 -10003 10013 -10014" universe="10000" />
<cell id="10010" material="10010" region="10000 -10001 10002 -10003 10014 -10015" universe="10000" />
<cell id="10011" material="10011" region="10000 -10001 10002 -10003 10015 -10016" universe="10000" />
<surface boundary="reflective" coeffs="0.0" id="10000" type="x-plane" />
<surface boundary="reflective" coeffs="10.0" id="10001" type="x-plane" />
<surface boundary="reflective" coeffs="0.0" id="10002" type="y-plane" />

12
tests/test_mg_basic/test_mg_basic.py
View file

@ -3,15 +3,11 @@
import os
import sys
sys.path.insert(0, os.pardir)
from testing_harness import TestHarness, PyAPITestHarness
import openmc
class MGBasicTestHarness(PyAPITestHarness):
def _build_inputs(self):
super(MGBasicTestHarness, self)._build_inputs()
from testing_harness import PyAPITestHarness
from openmc.examples import slab_mg
if __name__ == '__main__':
harness = MGBasicTestHarness('statepoint.10.*', False, mg=True)
model = slab_mg()
harness = PyAPITestHarness('statepoint.10.h5', model)
harness.main()

2
tests/test_mg_convert/test_mg_convert.py
View file

@ -206,5 +206,5 @@ class MGXSTestHarness(PyAPITestHarness):
if __name__ == '__main__':
harness = MGXSTestHarness('statepoint.10.*', False)
harness = MGXSTestHarness('statepoint.10.h5')
harness.main()

6
tests/test_mg_legendre/inputs_true.dat
View file

@ -1,8 +1,8 @@
<?xml version='1.0' encoding='utf-8'?>
<geometry>
<cell id="10000" material="10000" region="10000 -10001 10002 -10003 10004 -10005" universe="0" />
<cell id="10001" material="10001" region="10000 -10001 10002 -10003 10005 -10006" universe="0" />
<cell id="10002" material="10002" region="10000 -10001 10002 -10003 10006 -10007" universe="0" />
<cell id="10000" material="10000" region="10000 -10001 10002 -10003 10004 -10005" universe="10000" />
<cell id="10001" material="10001" region="10000 -10001 10002 -10003 10005 -10006" universe="10000" />
<cell id="10002" material="10002" region="10000 -10001 10002 -10003 10006 -10007" universe="10000" />
<surface boundary="reflective" coeffs="0.0" id="10000" type="x-plane" />
<surface boundary="reflective" coeffs="10.0" id="10001" type="x-plane" />
<surface boundary="reflective" coeffs="0.0" id="10002" type="y-plane" />

22
tests/test_mg_legendre/test_mg_legendre.py
View file

@ -5,24 +5,12 @@ import sys
sys.path.insert(0, os.pardir)
from testing_harness import PyAPITestHarness
from input_set import MGInputSet
class MGMaxOrderTestHarness(PyAPITestHarness):
def __init__(self, statepoint_name, tallies_present, mg=False):
PyAPITestHarness.__init__(self, statepoint_name, tallies_present)
self._input_set = MGInputSet()
def _build_inputs(self):
"""Write input XML files."""
reps = ['iso']
self._input_set.build_default_materials_and_geometry(reps=reps)
self._input_set.build_default_settings()
# Enforce Legendre scattering
self._input_set.settings.tabular_legendre = {'enable': False}
self._input_set.export()
from openmc.examples import slab_mg
if __name__ == '__main__':
harness = MGMaxOrderTestHarness('statepoint.10.*', False, mg=True)
model = slab_mg(reps=['iso'])
model.settings.tabular_legendre = {'enable': False}
harness = PyAPITestHarness('statepoint.10.h5', model)
harness.main()

6
tests/test_mg_max_order/inputs_true.dat
View file

@ -1,8 +1,8 @@
<?xml version='1.0' encoding='utf-8'?>
<geometry>
<cell id="10000" material="10000" region="10000 -10001 10002 -10003 10004 -10005" universe="0" />
<cell id="10001" material="10001" region="10000 -10001 10002 -10003 10005 -10006" universe="0" />
<cell id="10002" material="10002" region="10000 -10001 10002 -10003 10006 -10007" universe="0" />
<cell id="10000" material="10000" region="10000 -10001 10002 -10003 10004 -10005" universe="10000" />
<cell id="10001" material="10001" region="10000 -10001 10002 -10003 10005 -10006" universe="10000" />
<cell id="10002" material="10002" region="10000 -10001 10002 -10003 10006 -10007" universe="10000" />
<surface boundary="reflective" coeffs="0.0" id="10000" type="x-plane" />
<surface boundary="reflective" coeffs="10.0" id="10001" type="x-plane" />
<surface boundary="reflective" coeffs="0.0" id="10002" type="y-plane" />

21
tests/test_mg_max_order/test_mg_max_order.py
View file

@ -5,24 +5,11 @@ import sys
sys.path.insert(0, os.pardir)
from testing_harness import PyAPITestHarness
from input_set import MGInputSet
class MGMaxOrderTestHarness(PyAPITestHarness):
def __init__(self, statepoint_name, tallies_present, mg=False):
PyAPITestHarness.__init__(self, statepoint_name, tallies_present)
self._input_set = MGInputSet()
def _build_inputs(self):
"""Write input XML files."""
reps = ['iso']
self._input_set.build_default_materials_and_geometry(reps=reps)
self._input_set.build_default_settings()
# Set P1 scattering
self._input_set.settings.max_order = 1
self._input_set.export()
from openmc.examples import slab_mg
if __name__ == '__main__':
harness = MGMaxOrderTestHarness('statepoint.10.*', False, mg=True)
model = slab_mg(reps=['iso'])
model.settings.max_order = 1
harness = PyAPITestHarness('statepoint.10.h5', model)
harness.main()

24
tests/test_mg_nuclide/inputs_true.dat
View file

@ -1,17 +1,17 @@
<?xml version='1.0' encoding='utf-8'?>
<geometry>
<cell id="10000" material="10000" region="10000 -10001 10002 -10003 10004 -10005" universe="0" />
<cell id="10001" material="10001" region="10000 -10001 10002 -10003 10005 -10006" universe="0" />
<cell id="10002" material="10002" region="10000 -10001 10002 -10003 10006 -10007" universe="0" />
<cell id="10003" material="10003" region="10000 -10001 10002 -10003 10007 -10008" universe="0" />
<cell id="10004" material="10004" region="10000 -10001 10002 -10003 10008 -10009" universe="0" />
<cell id="10005" material="10005" region="10000 -10001 10002 -10003 10009 -10010" universe="0" />
<cell id="10006" material="10006" region="10000 -10001 10002 -10003 10010 -10011" universe="0" />
<cell id="10007" material="10007" region="10000 -10001 10002 -10003 10011 -10012" universe="0" />
<cell id="10008" material="10008" region="10000 -10001 10002 -10003 10012 -10013" universe="0" />
<cell id="10009" material="10009" region="10000 -10001 10002 -10003 10013 -10014" universe="0" />
<cell id="10010" material="10010" region="10000 -10001 10002 -10003 10014 -10015" universe="0" />
<cell id="10011" material="10011" region="10000 -10001 10002 -10003 10015 -10016" universe="0" />
<cell id="10000" material="10000" region="10000 -10001 10002 -10003 10004 -10005" universe="10000" />
<cell id="10001" material="10001" region="10000 -10001 10002 -10003 10005 -10006" universe="10000" />
<cell id="10002" material="10002" region="10000 -10001 10002 -10003 10006 -10007" universe="10000" />
<cell id="10003" material="10003" region="10000 -10001 10002 -10003 10007 -10008" universe="10000" />
<cell id="10004" material="10004" region="10000 -10001 10002 -10003 10008 -10009" universe="10000" />
<cell id="10005" material="10005" region="10000 -10001 10002 -10003 10009 -10010" universe="10000" />
<cell id="10006" material="10006" region="10000 -10001 10002 -10003 10010 -10011" universe="10000" />
<cell id="10007" material="10007" region="10000 -10001 10002 -10003 10011 -10012" universe="10000" />
<cell id="10008" material="10008" region="10000 -10001 10002 -10003 10012 -10013" universe="10000" />
<cell id="10009" material="10009" region="10000 -10001 10002 -10003 10013 -10014" universe="10000" />
<cell id="10010" material="10010" region="10000 -10001 10002 -10003 10014 -10015" universe="10000" />
<cell id="10011" material="10011" region="10000 -10001 10002 -10003 10015 -10016" universe="10000" />
<surface boundary="reflective" coeffs="0.0" id="10000" type="x-plane" />
<surface boundary="reflective" coeffs="10.0" id="10001" type="x-plane" />
<surface boundary="reflective" coeffs="0.0" id="10002" type="y-plane" />

17
tests/test_mg_nuclide/test_mg_nuclide.py
View file

@ -5,21 +5,10 @@ import sys
sys.path.insert(0, os.pardir)
from testing_harness import PyAPITestHarness
from input_set import MGInputSet
class MGNuclideTestHarness(PyAPITestHarness):
def __init__(self, statepoint_name, tallies_present, mg=False):
PyAPITestHarness.__init__(self, statepoint_name, tallies_present)
self._input_set = MGInputSet()
def _build_inputs(self):
"""Write input XML files."""
self._input_set.build_default_materials_and_geometry(as_macro=False)
self._input_set.build_default_settings()
self._input_set.export()
from openmc.examples import slab_mg
if __name__ == '__main__':
harness = MGNuclideTestHarness('statepoint.10.*', False, mg=True)
model = slab_mg(as_macro=False)
harness = PyAPITestHarness('statepoint.10.h5', model)
harness.main()

24
tests/test_mg_survival_biasing/inputs_true.dat
View file

@ -1,17 +1,17 @@
<?xml version='1.0' encoding='utf-8'?>
<geometry>
<cell id="10000" material="10000" region="10000 -10001 10002 -10003 10004 -10005" universe="0" />
<cell id="10001" material="10001" region="10000 -10001 10002 -10003 10005 -10006" universe="0" />
<cell id="10002" material="10002" region="10000 -10001 10002 -10003 10006 -10007" universe="0" />
<cell id="10003" material="10003" region="10000 -10001 10002 -10003 10007 -10008" universe="0" />
<cell id="10004" material="10004" region="10000 -10001 10002 -10003 10008 -10009" universe="0" />
<cell id="10005" material="10005" region="10000 -10001 10002 -10003 10009 -10010" universe="0" />
<cell id="10006" material="10006" region="10000 -10001 10002 -10003 10010 -10011" universe="0" />
<cell id="10007" material="10007" region="10000 -10001 10002 -10003 10011 -10012" universe="0" />
<cell id="10008" material="10008" region="10000 -10001 10002 -10003 10012 -10013" universe="0" />
<cell id="10009" material="10009" region="10000 -10001 10002 -10003 10013 -10014" universe="0" />
<cell id="10010" material="10010" region="10000 -10001 10002 -10003 10014 -10015" universe="0" />
<cell id="10011" material="10011" region="10000 -10001 10002 -10003 10015 -10016" universe="0" />
<cell id="10000" material="10000" region="10000 -10001 10002 -10003 10004 -10005" universe="10000" />
<cell id="10001" material="10001" region="10000 -10001 10002 -10003 10005 -10006" universe="10000" />
<cell id="10002" material="10002" region="10000 -10001 10002 -10003 10006 -10007" universe="10000" />
<cell id="10003" material="10003" region="10000 -10001 10002 -10003 10007 -10008" universe="10000" />
<cell id="10004" material="10004" region="10000 -10001 10002 -10003 10008 -10009" universe="10000" />
<cell id="10005" material="10005" region="10000 -10001 10002 -10003 10009 -10010" universe="10000" />
<cell id="10006" material="10006" region="10000 -10001 10002 -10003 10010 -10011" universe="10000" />
<cell id="10007" material="10007" region="10000 -10001 10002 -10003 10011 -10012" universe="10000" />
<cell id="10008" material="10008" region="10000 -10001 10002 -10003 10012 -10013" universe="10000" />
<cell id="10009" material="10009" region="10000 -10001 10002 -10003 10013 -10014" universe="10000" />
<cell id="10010" material="10010" region="10000 -10001 10002 -10003 10014 -10015" universe="10000" />
<cell id="10011" material="10011" region="10000 -10001 10002 -10003 10015 -10016" universe="10000" />
<surface boundary="reflective" coeffs="0.0" id="10000" type="x-plane" />
<surface boundary="reflective" coeffs="10.0" id="10001" type="x-plane" />
<surface boundary="reflective" coeffs="0.0" id="10002" type="y-plane" />

17
tests/test_mg_survival_biasing/test_mg_survival_biasing.py
View file

@ -3,19 +3,12 @@
import os
import sys
sys.path.insert(0, os.pardir)
from testing_harness import TestHarness, PyAPITestHarness
import openmc
class MGBasicTestHarness(PyAPITestHarness):
def _build_inputs(self):
"""Write input XML files."""
self._input_set.build_default_materials_and_geometry()
self._input_set.build_default_settings()
self._input_set.settings.survival_biasing = True
self._input_set.export()
from testing_harness import PyAPITestHarness
from openmc.examples import slab_mg
if __name__ == '__main__':
harness = MGBasicTestHarness('statepoint.10.h5', False, mg=True)
model = slab_mg()
model.settings.survival_biasing = True
harness = PyAPITestHarness('statepoint.10.h5', model)
harness.main()

24
tests/test_mg_tallies/inputs_true.dat
View file

@ -1,17 +1,17 @@
<?xml version='1.0' encoding='utf-8'?>
<geometry>
<cell id="10000" material="10000" region="10000 -10001 10002 -10003 10004 -10005" universe="0" />
<cell id="10001" material="10001" region="10000 -10001 10002 -10003 10005 -10006" universe="0" />
<cell id="10002" material="10002" region="10000 -10001 10002 -10003 10006 -10007" universe="0" />
<cell id="10003" material="10003" region="10000 -10001 10002 -10003 10007 -10008" universe="0" />
<cell id="10004" material="10004" region="10000 -10001 10002 -10003 10008 -10009" universe="0" />
<cell id="10005" material="10005" region="10000 -10001 10002 -10003 10009 -10010" universe="0" />
<cell id="10006" material="10006" region="10000 -10001 10002 -10003 10010 -10011" universe="0" />
<cell id="10007" material="10007" region="10000 -10001 10002 -10003 10011 -10012" universe="0" />
<cell id="10008" material="10008" region="10000 -10001 10002 -10003 10012 -10013" universe="0" />
<cell id="10009" material="10009" region="10000 -10001 10002 -10003 10013 -10014" universe="0" />
<cell id="10010" material="10010" region="10000 -10001 10002 -10003 10014 -10015" universe="0" />
<cell id="10011" material="10011" region="10000 -10001 10002 -10003 10015 -10016" universe="0" />
<cell id="10000" material="10000" region="10000 -10001 10002 -10003 10004 -10005" universe="10000" />
<cell id="10001" material="10001" region="10000 -10001 10002 -10003 10005 -10006" universe="10000" />
<cell id="10002" material="10002" region="10000 -10001 10002 -10003 10006 -10007" universe="10000" />
<cell id="10003" material="10003" region="10000 -10001 10002 -10003 10007 -10008" universe="10000" />
<cell id="10004" material="10004" region="10000 -10001 10002 -10003 10008 -10009" universe="10000" />
<cell id="10005" material="10005" region="10000 -10001 10002 -10003 10009 -10010" universe="10000" />
<cell id="10006" material="10006" region="10000 -10001 10002 -10003 10010 -10011" universe="10000" />
<cell id="10007" material="10007" region="10000 -10001 10002 -10003 10011 -10012" universe="10000" />
<cell id="10008" material="10008" region="10000 -10001 10002 -10003 10012 -10013" universe="10000" />
<cell id="10009" material="10009" region="10000 -10001 10002 -10003 10013 -10014" universe="10000" />
<cell id="10010" material="10010" region="10000 -10001 10002 -10003 10014 -10015" universe="10000" />
<cell id="10011" material="10011" region="10000 -10001 10002 -10003 10015 -10016" universe="10000" />
<surface boundary="reflective" coeffs="0.0" id="10000" type="x-plane" />
<surface boundary="reflective" coeffs="10.0" id="10001" type="x-plane" />
<surface boundary="reflective" coeffs="0.0" id="10002" type="y-plane" />

178
tests/test_mg_tallies/test_mg_tallies.py
View file

@ -5,100 +5,92 @@ import sys
sys.path.insert(0, os.pardir)
from testing_harness import HashedPyAPITestHarness
import openmc
class MGTalliesTestHarness(HashedPyAPITestHarness):
def _build_inputs(self):
"""Write input XML files."""
self._input_set.build_default_materials_and_geometry(as_macro=False)
self._input_set.build_default_settings()
# Instantiate a tally mesh
mesh = openmc.Mesh(mesh_id=1)
mesh.type = 'regular'
mesh.dimension = [1, 1, 10]
mesh.lower_left = [0.0, 0.0, 0.0]
mesh.upper_right = [10, 10, 5]
# Instantiate some tally filters
energy_filter = openmc.EnergyFilter([0.0, 20.0e6])
energyout_filter = openmc.EnergyoutFilter([0.0, 20.0e6])
matching_energy_filter = openmc.EnergyFilter([1e-5, 0.0635, 10.0,
1.0e2, 1.0e3, 0.5e6,
1.0e6, 20.0e6])
matching_eout_filter = openmc.EnergyoutFilter([1e-5, 0.0635, 10.0,
1.0e2, 1.0e3, 0.5e6,
1.0e6, 20.0e6])
mesh_filter = openmc.MeshFilter(mesh)
mat_ids = [mat.id for mat in self._input_set.materials]
mat_filter = openmc.MaterialFilter(mat_ids)
nuclides = [xs.name for xs in self._input_set.xs_data]
scores= {False: ['total', 'absorption', 'flux', 'fission', 'nu-fission'],
True: ['total', 'absorption', 'fission', 'nu-fission']}
tallies = []
for do_nuclides in [False, True]:
tallies.append(openmc.Tally())
tallies[-1].filters = [mesh_filter]
tallies[-1].estimator = 'analog'
tallies[-1].scores = scores[do_nuclides]
if do_nuclides:
tallies[-1].nuclides = nuclides
tallies.append(openmc.Tally())
tallies[-1].filters = [mesh_filter]
tallies[-1].estimator = 'tracklength'
tallies[-1].scores = scores[do_nuclides]
if do_nuclides:
tallies[-1].nuclides = nuclides
# Impose energy bins that dont match the MG structure and those
# that do
for match_energy_bins in [False, True]:
if match_energy_bins:
e_filter = matching_energy_filter
eout_filter = matching_eout_filter
else:
e_filter = energy_filter
eout_filter = energyout_filter
tallies.append(openmc.Tally())
tallies[-1].filters = [mat_filter, e_filter]
tallies[-1].estimator = 'analog'
tallies[-1].scores = scores[do_nuclides] + ['scatter',
'nu-scatter']
if do_nuclides:
tallies[-1].nuclides = nuclides
tallies.append(openmc.Tally())
tallies[-1].filters = [mat_filter, e_filter]
tallies[-1].estimator = 'collision'
tallies[-1].scores = scores[do_nuclides]
if do_nuclides:
tallies[-1].nuclides = nuclides
tallies.append(openmc.Tally())
tallies[-1].filters = [mat_filter, e_filter]
tallies[-1].estimator = 'tracklength'
tallies[-1].scores = scores[do_nuclides]
if do_nuclides:
tallies[-1].nuclides = nuclides
tallies.append(openmc.Tally())
tallies[-1].filters = [mat_filter, e_filter, eout_filter]
tallies[-1].scores = ['scatter', 'nu-scatter', 'nu-fission']
if do_nuclides:
tallies[-1].nuclides = nuclides
self._input_set.tallies = openmc.Tallies(tallies)
self._input_set.export()
from openmc.examples import slab_mg
if __name__ == '__main__':
harness = MGTalliesTestHarness('statepoint.10.h5', True, mg=True)
model = slab_mg(as_macro=False)
# Instantiate a tally mesh
mesh = openmc.Mesh(mesh_id=1)
mesh.type = 'regular'
mesh.dimension = [1, 1, 10]
mesh.lower_left = [0.0, 0.0, 0.0]
mesh.upper_right = [10, 10, 5]
# Instantiate some tally filters
energy_filter = openmc.EnergyFilter([0.0, 20.0e6])
energyout_filter = openmc.EnergyoutFilter([0.0, 20.0e6])
energies = [1e-5, 0.0635, 10.0, 1.0e2, 1.0e3, 0.5e6, 1.0e6, 20.0e6]
matching_energy_filter = openmc.EnergyFilter(energies)
matching_eout_filter = openmc.EnergyoutFilter(energies)
mesh_filter = openmc.MeshFilter(mesh)
mat_ids = [mat.id for mat in model.materials]
mat_filter = openmc.MaterialFilter(mat_ids)
nuclides = [xs.name for xs in model.xs_data]
scores= {False: ['total', 'absorption', 'flux', 'fission', 'nu-fission'],
True: ['total', 'absorption', 'fission', 'nu-fission']}
for do_nuclides in [False, True]:
t = openmc.Tally()
t.filters = [mesh_filter]
t.estimator = 'analog'
t.scores = scores[do_nuclides]
if do_nuclides:
t.nuclides = nuclides
model.tallies.append(t)
t = openmc.Tally()
t.filters = [mesh_filter]
t.estimator = 'tracklength'
t.scores = scores[do_nuclides]
if do_nuclides:
t.nuclides = nuclides
model.tallies.append(t)
# Impose energy bins that dont match the MG structure and those
# that do
for match_energy_bins in [False, True]:
if match_energy_bins:
e_filter = matching_energy_filter
eout_filter = matching_eout_filter
else:
e_filter = energy_filter
eout_filter = energyout_filter
t = openmc.Tally()
t.filters = [mat_filter, e_filter]
t.estimator = 'analog'
t.scores = scores[do_nuclides] + ['scatter', 'nu-scatter']
if do_nuclides:
t.nuclides = nuclides
model.tallies.append(t)
t = openmc.Tally()
t.filters = [mat_filter, e_filter]
t.estimator = 'collision'
t.scores = scores[do_nuclides]
if do_nuclides:
t.nuclides = nuclides
model.tallies.append(t)
t = openmc.Tally()
t.filters = [mat_filter, e_filter]
t.estimator = 'tracklength'
t.scores = scores[do_nuclides]
if do_nuclides:
t.nuclides = nuclides
model.tallies.append(t)
t = openmc.Tally()
t.filters = [mat_filter, e_filter, eout_filter]
t.scores = ['scatter', 'nu-scatter', 'nu-fission']
if do_nuclides:
t.nuclides = nuclides
model.tallies.append(t)
harness = HashedPyAPITestHarness('statepoint.10.h5', model)
harness.main()

10
tests/test_mgxs_library_ce_to_mg/inputs_true.dat
View file

@ -1,8 +1,8 @@
<?xml version='1.0' encoding='utf-8'?>
<geometry>
<cell id="10000" material="10000" name="cell 1" region="-10000" universe="0" />
<cell id="10001" material="10001" name="cell 3" region="10000 -10001" universe="0" />
<cell id="10002" material="10002" name="cell 2" region="10001 10002 -10003 10004 -10005" universe="0" />
<cell id="10000" material="10000" name="Fuel" region="-10000" universe="0" />
<cell id="10001" material="10001" name="Cladding" region="10000 -10001" universe="0" />
<cell id="10002" material="10002" name="Water" region="10001 10002 -10003 10004 -10005" universe="0" />
<surface coeffs="0 0 0.39218" id="10000" name="Fuel OR" type="z-cylinder" />
<surface coeffs="0 0 0.4572" id="10001" name="Clad OR" type="z-cylinder" />
<surface boundary="reflective" coeffs="-0.63" id="10002" name="left" type="x-plane" />
@ -12,14 +12,14 @@
</geometry>
<?xml version='1.0' encoding='utf-8'?>
<materials>
<material id="10000" name="Fuel">
<material id="10000" name="UO2 (2.4%)">
<density units="g/cm3" value="10.29769" />
<nuclide ao="4.4843e-06" name="U234" />
<nuclide ao="0.00055815" name="U235" />
<nuclide ao="0.022408" name="U238" />
<nuclide ao="0.045829" name="O16" />
</material>
<material id="10001" name="Cladding">
<material id="10001" name="Zircaloy">
<density units="g/cm3" value="6.55" />
<nuclide ao="0.021827" name="Zr90" />
<nuclide ao="0.00476" name="Zr91" />

37
tests/test_mgxs_library_ce_to_mg/test_mgxs_library_ce_to_mg.py
View file

@ -3,27 +3,23 @@
import os
import sys
import glob
import hashlib
sys.path.insert(0, os.pardir)
from testing_harness import PyAPITestHarness
from input_set import PinCellInputSet
import openmc
import openmc.mgxs
from openmc.examples import pwr_pin_cell
class MGXSTestHarness(PyAPITestHarness):
def _build_inputs(self):
# Set the input set to use the pincell model
self._input_set = PinCellInputSet()
def __init__(self, *args, **kwargs):
# Generate inputs using parent class routine
super(MGXSTestHarness, self)._build_inputs()
super(MGXSTestHarness, self).__init__(*args, **kwargs)
# Initialize a two-group structure
energy_groups = openmc.mgxs.EnergyGroups(group_edges=[0, 0.625, 20.e6])
# Initialize MGXS Library for a few cross section types
self.mgxs_lib = openmc.mgxs.Library(self._input_set.geometry)
self.mgxs_lib = openmc.mgxs.Library(self._model.geometry)
self.mgxs_lib.by_nuclide = False
self.mgxs_lib.mgxs_types = ['total', 'absorption', 'nu-fission matrix',
'nu-scatter matrix', 'multiplicity matrix']
@ -34,9 +30,7 @@ class MGXSTestHarness(PyAPITestHarness):
self.mgxs_lib.build_library()
# Initialize a tallies file
self._input_set.tallies = openmc.Tallies()
self.mgxs_lib.add_to_tallies_file(self._input_set.tallies, merge=False)
self._input_set.tallies.export_to_xml()
self.mgxs_lib.add_to_tallies_file(self._model.tallies, merge=False)
def _run_openmc(self):
# Initial run
@ -55,18 +49,18 @@ class MGXSTestHarness(PyAPITestHarness):
statepoint = glob.glob(os.path.join(os.getcwd(), self._sp_name))[0]
sp = openmc.StatePoint(statepoint)
self.mgxs_lib.load_from_statepoint(sp)
self._input_set.mgxs_file, self._input_set.materials, \
self._input_set.geometry = self.mgxs_lib.create_mg_mode()
self._model.mgxs_file, self._model.materials, \
self._model.geometry = self.mgxs_lib.create_mg_mode()
# Modify materials and settings so we can run in MG mode
self._input_set.materials.cross_sections = './mgxs.h5'
self._input_set.settings.energy_mode = 'multi-group'
self._model.materials.cross_sections = './mgxs.h5'
self._model.settings.energy_mode = 'multi-group'
# Write modified input files
self._input_set.settings.export_to_xml()
self._input_set.geometry.export_to_xml()
self._input_set.materials.export_to_xml()
self._input_set.mgxs_file.export_to_hdf5()
self._model.settings.export_to_xml()
self._model.geometry.export_to_xml()
self._model.materials.export_to_xml()
self._model.mgxs_file.export_to_hdf5()
# Dont need tallies.xml, so remove the file
if os.path.exists('./tallies.xml'):
os.remove('./tallies.xml')
@ -92,5 +86,8 @@ class MGXSTestHarness(PyAPITestHarness):
if __name__ == '__main__':
harness = MGXSTestHarness('statepoint.10.*', False)
# Set the input set to use the pincell model
model = pwr_pin_cell()
harness = MGXSTestHarness('statepoint.10.h5', model)
harness.main()

10
tests/test_mgxs_library_condense/inputs_true.dat
View file

@ -1,8 +1,8 @@
<?xml version='1.0' encoding='utf-8'?>
<geometry>
<cell id="10000" material="10000" name="cell 1" region="-10000" universe="0" />
<cell id="10001" material="10001" name="cell 3" region="10000 -10001" universe="0" />
<cell id="10002" material="10002" name="cell 2" region="10001 10002 -10003 10004 -10005" universe="0" />
<cell id="10000" material="10000" name="Fuel" region="-10000" universe="0" />
<cell id="10001" material="10001" name="Cladding" region="10000 -10001" universe="0" />
<cell id="10002" material="10002" name="Water" region="10001 10002 -10003 10004 -10005" universe="0" />
<surface coeffs="0 0 0.39218" id="10000" name="Fuel OR" type="z-cylinder" />
<surface coeffs="0 0 0.4572" id="10001" name="Clad OR" type="z-cylinder" />
<surface boundary="reflective" coeffs="-0.63" id="10002" name="left" type="x-plane" />
@ -12,14 +12,14 @@
</geometry>
<?xml version='1.0' encoding='utf-8'?>
<materials>
<material id="10000" name="Fuel">
<material id="10000" name="UO2 (2.4%)">
<density units="g/cm3" value="10.29769" />
<nuclide ao="4.4843e-06" name="U234" />
<nuclide ao="0.00055815" name="U235" />
<nuclide ao="0.022408" name="U238" />
<nuclide ao="0.045829" name="O16" />
</material>
<material id="10001" name="Cladding">
<material id="10001" name="Zircaloy">
<density units="g/cm3" value="6.55" />
<nuclide ao="0.021827" name="Zr90" />
<nuclide ao="0.00476" name="Zr91" />

22
tests/test_mgxs_library_condense/test_mgxs_library_condense.py
View file

@ -6,24 +6,20 @@ import glob
import hashlib
sys.path.insert(0, os.pardir)
from testing_harness import PyAPITestHarness
from input_set import PinCellInputSet
import openmc
import openmc.mgxs
from openmc.examples import pwr_pin_cell
class MGXSTestHarness(PyAPITestHarness):
def _build_inputs(self):
# Set the input set to use the pincell model
self._input_set = PinCellInputSet()
# Generate inputs using parent class routine
super(MGXSTestHarness, self)._build_inputs()
def __init__(self, *args, **kwargs):
super(MGXSTestHarness, self).__init__(*args, **kwargs)
# Initialize a two-group structure
energy_groups = openmc.mgxs.EnergyGroups(group_edges=[0, 0.625, 20.e6])
# Initialize MGXS Library for a few cross section types
self.mgxs_lib = openmc.mgxs.Library(self._input_set.geometry)
self.mgxs_lib = openmc.mgxs.Library(self._model.geometry)
self.mgxs_lib.by_nuclide = False
# Test all MGXS types
@ -35,10 +31,8 @@ class MGXSTestHarness(PyAPITestHarness):
self.mgxs_lib.domain_type = 'material'
self.mgxs_lib.build_library()
# Initialize a tallies file
self._input_set.tallies = openmc.Tallies()
self.mgxs_lib.add_to_tallies_file(self._input_set.tallies, merge=False)
self._input_set.tallies.export_to_xml()
# Add tallies
self.mgxs_lib.add_to_tallies_file(self._model.tallies, merge=False)
def _get_results(self, hash_output=False):
"""Digest info in the statepoint and return as a string."""
@ -72,5 +66,7 @@ class MGXSTestHarness(PyAPITestHarness):
if __name__ == '__main__':
harness = MGXSTestHarness('statepoint.10.*', True)
# Use the pincell model
model = pwr_pin_cell()
harness = MGXSTestHarness('statepoint.10.h5', model)
harness.main()

2
tests/test_mgxs_library_distribcell/inputs_true.dat
View file

@ -6,7 +6,7 @@
<cell id="10003" material="10002" name="guide tube inner water" region="-10000" universe="10001" />
<cell id="10004" material="10001" name="guide tube clad" region="10000 -10001" universe="10001" />
<cell id="10005" material="10002" name="guide tube outer water" region="10001" universe="10001" />
<cell fill="10002" id="10006" name="root cell" region="10002 -10003 10004 -10005" universe="0" />
<cell fill="10002" id="10006" name="root cell" region="10002 -10003 10004 -10005" universe="10003" />
<lattice id="10002" name="Fuel Assembly">
<pitch>1.26 1.26</pitch>
<dimension>17 17</dimension>

21
tests/test_mgxs_library_distribcell/test_mgxs_library_distribcell.py
View file

@ -6,25 +6,22 @@ import glob
import hashlib
sys.path.insert(0, os.pardir)
from testing_harness import PyAPITestHarness
from input_set import AssemblyInputSet
import openmc
import openmc.mgxs
from openmc.examples import pwr_assembly
class MGXSTestHarness(PyAPITestHarness):
def _build_inputs(self):
# Set the input set to use the pincell model
self._input_set = AssemblyInputSet()
def __init__(self, *args, **kwargs):
# Generate inputs using parent class routine
super(MGXSTestHarness, self)._build_inputs()
super(MGXSTestHarness, self).__init__(*args, **kwargs)
# Initialize a one-group structure
energy_groups = openmc.mgxs.EnergyGroups(group_edges=[0, 20.e6])
# Initialize MGXS Library for a few cross section types
# for one material-filled cell in the geometry
self.mgxs_lib = openmc.mgxs.Library(self._input_set.geometry)
self.mgxs_lib = openmc.mgxs.Library(self._model.geometry)
self.mgxs_lib.by_nuclide = False
# Test all MGXS types
@ -38,10 +35,9 @@ class MGXSTestHarness(PyAPITestHarness):
self.mgxs_lib.domains = [c for c in cells if c.name == 'fuel']
self.mgxs_lib.build_library()
# Initialize a tallies file
self._input_set.tallies = openmc.Tallies()
self.mgxs_lib.add_to_tallies_file(self._input_set.tallies, merge=False)
self._input_set.tallies.export_to_xml()
# Add tallies
self.mgxs_lib.add_to_tallies_file(self._model.tallies, merge=False)
self._model.tallies.export_to_xml()
def _get_results(self, hash_output=False):
"""Digest info in the statepoint and return as a string."""
@ -74,5 +70,6 @@ class MGXSTestHarness(PyAPITestHarness):
if __name__ == '__main__':
harness = MGXSTestHarness('statepoint.10.h5', True)
model = pwr_assembly()
harness = MGXSTestHarness('statepoint.10.h5', model)
harness.main()

10
tests/test_mgxs_library_hdf5/inputs_true.dat
View file

@ -1,8 +1,8 @@
<?xml version='1.0' encoding='utf-8'?>
<geometry>
<cell id="10000" material="10000" name="cell 1" region="-10000" universe="0" />
<cell id="10001" material="10001" name="cell 3" region="10000 -10001" universe="0" />
<cell id="10002" material="10002" name="cell 2" region="10001 10002 -10003 10004 -10005" universe="0" />
<cell id="10000" material="10000" name="Fuel" region="-10000" universe="0" />
<cell id="10001" material="10001" name="Cladding" region="10000 -10001" universe="0" />
<cell id="10002" material="10002" name="Water" region="10001 10002 -10003 10004 -10005" universe="0" />
<surface coeffs="0 0 0.39218" id="10000" name="Fuel OR" type="z-cylinder" />
<surface coeffs="0 0 0.4572" id="10001" name="Clad OR" type="z-cylinder" />
<surface boundary="reflective" coeffs="-0.63" id="10002" name="left" type="x-plane" />
@ -12,14 +12,14 @@
</geometry>
<?xml version='1.0' encoding='utf-8'?>
<materials>
<material id="10000" name="Fuel">
<material id="10000" name="UO2 (2.4%)">
<density units="g/cm3" value="10.29769" />
<nuclide ao="4.4843e-06" name="U234" />
<nuclide ao="0.00055815" name="U235" />
<nuclide ao="0.022408" name="U238" />
<nuclide ao="0.045829" name="O16" />
</material>
<material id="10001" name="Cladding">
<material id="10001" name="Zircaloy">
<density units="g/cm3" value="6.55" />
<nuclide ao="0.021827" name="Zr90" />
<nuclide ao="0.00476" name="Zr91" />

48
tests/test_mgxs_library_hdf5/test_mgxs_library_hdf5.py
View file

@ -10,27 +10,24 @@ import h5py
sys.path.insert(0, os.pardir)
from testing_harness import PyAPITestHarness
from input_set import PinCellInputSet
import openmc
import openmc.mgxs
from openmc.examples import pwr_pin_cell
np.set_printoptions(formatter={'float_kind': '{:.8e}'.format})
class MGXSTestHarness(PyAPITestHarness):
def _build_inputs(self):
# Set the input set to use the pincell model
self._input_set = PinCellInputSet()
def __init__(self, *args, **kwargs):
# Generate inputs using parent class routine
super(MGXSTestHarness, self)._build_inputs()
super(MGXSTestHarness, self).__init__(*args, **kwargs)
# Initialize a two-group structure
energy_groups = openmc.mgxs.EnergyGroups(group_edges=[0, 0.625, 20.e6])
# Initialize MGXS Library for a few cross section types
self.mgxs_lib = openmc.mgxs.Library(self._input_set.geometry)
self.mgxs_lib = openmc.mgxs.Library(self._model.geometry)
self.mgxs_lib.by_nuclide = False
# Test all MGXS types
@ -42,10 +39,8 @@ class MGXSTestHarness(PyAPITestHarness):
self.mgxs_lib.domain_type = 'material'
self.mgxs_lib.build_library()
# Initialize a tallies file
self._input_set.tallies = openmc.Tallies()
self.mgxs_lib.add_to_tallies_file(self._input_set.tallies, merge=False)
self._input_set.tallies.export_to_xml()
# Add tallies
self.mgxs_lib.add_to_tallies_file(self._model.tallies, merge=False)
def _get_results(self, hash_output=False):
"""Digest info in the statepoint and return as a string."""
@ -59,21 +54,18 @@ class MGXSTestHarness(PyAPITestHarness):
# Export the MGXS Library to an HDF5 file
self.mgxs_lib.build_hdf5_store(directory='.')
# Open the MGXS HDF5 file
f = h5py.File('mgxs.h5', 'r')
with h5py.File('mgxs.h5', 'r') as f:
# Build a string from the datasets in the HDF5 file
outstr = ''
for domain in self.mgxs_lib.domains:
for mgxs_type in self.mgxs_lib.mgxs_types:
outstr += 'domain={0} type={1}\n'.format(domain.id, mgxs_type)
avg_key = 'material/{0}/{1}/average'.format(domain.id, mgxs_type)
std_key = 'material/{0}/{1}/std. dev.'.format(domain.id, mgxs_type)
outstr += '{}\n{}\n'.format(f[avg_key][...], f[std_key][...])
# Close the MGXS HDF5 file
f.close()
# Build a string from the datasets in the HDF5 file
outstr = ''
for domain in self.mgxs_lib.domains:
for mgxs_type in self.mgxs_lib.mgxs_types:
outstr += 'domain={0} type={1}\n'.format(domain.id, mgxs_type)
avg_key = 'material/{0}/{1}/average'.format(domain.id, mgxs_type)
std_key = 'material/{0}/{1}/std. dev.'.format(domain.id, mgxs_type)
outstr += '{}\n{}\n'.format(f[avg_key][...], f[std_key][...])
# Hash the results if necessary
if hash_output:
@ -85,12 +77,12 @@ class MGXSTestHarness(PyAPITestHarness):
def _cleanup(self):
super(MGXSTestHarness, self)._cleanup()
f = os.path.join(os.getcwd(), 'tallies.xml')
if os.path.exists(f): os.remove(f)
f = os.path.join(os.getcwd(), 'mgxs.h5')
if os.path.exists(f): os.remove(f)
if os.path.exists(f):
os.remove(f)
if __name__ == '__main__':
harness = MGXSTestHarness('statepoint.10.*', True)
model = pwr_pin_cell()
harness = MGXSTestHarness('statepoint.10.h5', model)
harness.main()

6
tests/test_mgxs_library_mesh/inputs_true.dat
View file

@ -148,7 +148,7 @@
</geometry>
<?xml version='1.0' encoding='utf-8'?>
<materials>
<material id="1" name="Fuel">
<material id="1" name="UOX fuel">
<density units="g/cm3" value="10.062" />
<nuclide ao="4.9476e-06" name="U234" />
<nuclide ao="0.00048218" name="U235" />
@ -197,7 +197,7 @@
<nuclide name="Zr96" wo="0.0216912612" />
<sab name="c_H_in_H2O" />
</material>
<material id="2" name="Cladding">
<material id="2" name="Zircaloy">
<density units="g/cm3" value="5.77" />
<nuclide ao="0.5145" name="Zr90" />
<nuclide ao="0.1122" name="Zr91" />
@ -249,7 +249,7 @@
<nuclide name="Cr52" wo="0.145407678031" />
<sab name="c_H_in_H2O" />
</material>
<material id="7" name="Upper radial reflector /Top plate region">
<material id="7" name="Upper radial reflector / Top plate region">
<density units="g/cm3" value="4.28" />
<nuclide name="H1" wo="0.0086117" />
<nuclide name="O16" wo="0.0683369" />

15
tests/test_mgxs_library_mesh/test_mgxs_library_mesh.py
View file

@ -11,16 +11,15 @@ import openmc.mgxs
class MGXSTestHarness(PyAPITestHarness):
def _build_inputs(self):
# Generate inputs using parent class routine
super(MGXSTestHarness, self)._build_inputs()
def __init__(self, *args, **kwargs):
super(MGXSTestHarness, self).__init__(*args, **kwargs)
# Initialize a one-group structure
energy_groups = openmc.mgxs.EnergyGroups(group_edges=[0, 20.e6])
# Initialize MGXS Library for a few cross section types
# for one material-filled cell in the geometry
self.mgxs_lib = openmc.mgxs.Library(self._input_set.geometry)
self.mgxs_lib = openmc.mgxs.Library(self._model.geometry)
self.mgxs_lib.by_nuclide = False
# Test all MGXS types
@ -41,10 +40,8 @@ class MGXSTestHarness(PyAPITestHarness):
self.mgxs_lib.domains = [mesh]
self.mgxs_lib.build_library()
# Initialize a tallies file
self._input_set.tallies = openmc.Tallies()
self.mgxs_lib.add_to_tallies_file(self._input_set.tallies, merge=False)
self._input_set.tallies.export_to_xml()
# Add tallies
self.mgxs_lib.add_to_tallies_file(self._model.tallies, merge=False)
def _get_results(self, hash_output=False):
"""Digest info in the statepoint and return as a string."""
@ -74,5 +71,5 @@ class MGXSTestHarness(PyAPITestHarness):
if __name__ == '__main__':
harness = MGXSTestHarness('statepoint.10.h5', True)
harness = MGXSTestHarness('statepoint.10.h5')
harness.main()

10
tests/test_mgxs_library_no_nuclides/inputs_true.dat
View file

@ -1,8 +1,8 @@
<?xml version='1.0' encoding='utf-8'?>
<geometry>
<cell id="10000" material="10000" name="cell 1" region="-10000" universe="0" />
<cell id="10001" material="10001" name="cell 3" region="10000 -10001" universe="0" />
<cell id="10002" material="10002" name="cell 2" region="10001 10002 -10003 10004 -10005" universe="0" />
<cell id="10000" material="10000" name="Fuel" region="-10000" universe="0" />
<cell id="10001" material="10001" name="Cladding" region="10000 -10001" universe="0" />
<cell id="10002" material="10002" name="Water" region="10001 10002 -10003 10004 -10005" universe="0" />
<surface coeffs="0 0 0.39218" id="10000" name="Fuel OR" type="z-cylinder" />
<surface coeffs="0 0 0.4572" id="10001" name="Clad OR" type="z-cylinder" />
<surface boundary="reflective" coeffs="-0.63" id="10002" name="left" type="x-plane" />
@ -12,14 +12,14 @@
</geometry>
<?xml version='1.0' encoding='utf-8'?>
<materials>
<material id="10000" name="Fuel">
<material id="10000" name="UO2 (2.4%)">
<density units="g/cm3" value="10.29769" />
<nuclide ao="4.4843e-06" name="U234" />
<nuclide ao="0.00055815" name="U235" />
<nuclide ao="0.022408" name="U238" />
<nuclide ao="0.045829" name="O16" />
</material>
<material id="10001" name="Cladding">
<material id="10001" name="Zircaloy">
<density units="g/cm3" value="6.55" />
<nuclide ao="0.021827" name="Zr90" />
<nuclide ao="0.00476" name="Zr91" />

20
tests/test_mgxs_library_no_nuclides/test_mgxs_library_no_nuclides.py
View file

@ -6,24 +6,21 @@ import glob
import hashlib
sys.path.insert(0, os.pardir)
from testing_harness import PyAPITestHarness
from input_set import PinCellInputSet
import openmc
import openmc.mgxs
from openmc.examples import pwr_pin_cell
class MGXSTestHarness(PyAPITestHarness):
def _build_inputs(self):
# Set the input set to use the pincell model
self._input_set = PinCellInputSet()
def __init__(self, *args, **kwargs):
# Generate inputs using parent class routine
super(MGXSTestHarness, self)._build_inputs()
super(MGXSTestHarness, self).__init__(*args, **kwargs)
# Initialize a two-group structure
energy_groups = openmc.mgxs.EnergyGroups(group_edges=[0, 0.625, 20.e6])
# Initialize MGXS Library for a few cross section types
self.mgxs_lib = openmc.mgxs.Library(self._input_set.geometry)
self.mgxs_lib = openmc.mgxs.Library(self._model.geometry)
self.mgxs_lib.by_nuclide = False
# Test all MGXS types
@ -35,10 +32,8 @@ class MGXSTestHarness(PyAPITestHarness):
self.mgxs_lib.domain_type = 'material'
self.mgxs_lib.build_library()
# Initialize a tallies file
self._input_set.tallies = openmc.Tallies()
self.mgxs_lib.add_to_tallies_file(self._input_set.tallies, merge=False)
self._input_set.tallies.export_to_xml()
# Add tallies
self.mgxs_lib.add_to_tallies_file(self._model.tallies, merge=False)
def _get_results(self, hash_output=False):
"""Digest info in the statepoint and return as a string."""
@ -68,5 +63,6 @@ class MGXSTestHarness(PyAPITestHarness):
if __name__ == '__main__':
harness = MGXSTestHarness('statepoint.10.h5', True)
model = pwr_pin_cell()
harness = MGXSTestHarness('statepoint.10.h5', model)
harness.main()

10
tests/test_mgxs_library_nuclides/inputs_true.dat
View file

@ -1,8 +1,8 @@
<?xml version='1.0' encoding='utf-8'?>
<geometry>
<cell id="10000" material="10000" name="cell 1" region="-10000" universe="0" />
<cell id="10001" material="10001" name="cell 3" region="10000 -10001" universe="0" />
<cell id="10002" material="10002" name="cell 2" region="10001 10002 -10003 10004 -10005" universe="0" />
<cell id="10000" material="10000" name="Fuel" region="-10000" universe="0" />
<cell id="10001" material="10001" name="Cladding" region="10000 -10001" universe="0" />
<cell id="10002" material="10002" name="Water" region="10001 10002 -10003 10004 -10005" universe="0" />
<surface coeffs="0 0 0.39218" id="10000" name="Fuel OR" type="z-cylinder" />
<surface coeffs="0 0 0.4572" id="10001" name="Clad OR" type="z-cylinder" />
<surface boundary="reflective" coeffs="-0.63" id="10002" name="left" type="x-plane" />
@ -12,14 +12,14 @@
</geometry>
<?xml version='1.0' encoding='utf-8'?>
<materials>
<material id="10000" name="Fuel">
<material id="10000" name="UO2 (2.4%)">
<density units="g/cm3" value="10.29769" />
<nuclide ao="4.4843e-06" name="U234" />
<nuclide ao="0.00055815" name="U235" />
<nuclide ao="0.022408" name="U238" />
<nuclide ao="0.045829" name="O16" />
</material>
<material id="10001" name="Cladding">
<material id="10001" name="Zircaloy">
<density units="g/cm3" value="6.55" />
<nuclide ao="0.021827" name="Zr90" />
<nuclide ao="0.00476" name="Zr91" />

21
tests/test_mgxs_library_nuclides/test_mgxs_library_nuclides.py
View file

@ -6,24 +6,20 @@ import glob
import hashlib
sys.path.insert(0, os.pardir)
from testing_harness import PyAPITestHarness
from input_set import PinCellInputSet
import openmc
import openmc.mgxs
from openmc.examples import pwr_pin_cell
class MGXSTestHarness(PyAPITestHarness):
def _build_inputs(self):
# Set the input set to use the pincell model
self._input_set = PinCellInputSet()
# Generate inputs using parent class routine
super(MGXSTestHarness, self)._build_inputs()
def __init__(self, *args, **kwargs):
super(MGXSTestHarness, self).__init__(*args, **kwargs)
# Initialize a two-group structure
energy_groups = openmc.mgxs.EnergyGroups(group_edges=[0, 0.625, 20.e6])
# Initialize MGXS Library for a few cross section types
self.mgxs_lib = openmc.mgxs.Library(self._input_set.geometry)
self.mgxs_lib = openmc.mgxs.Library(self._model.geometry)
self.mgxs_lib.by_nuclide = True
# Test all MGXS types
self.mgxs_lib.mgxs_types = openmc.mgxs.MGXS_TYPES
@ -32,10 +28,8 @@ class MGXSTestHarness(PyAPITestHarness):
self.mgxs_lib.domain_type = 'material'
self.mgxs_lib.build_library()
# Initialize a tallies file
self._input_set.tallies = openmc.Tallies()
self.mgxs_lib.add_to_tallies_file(self._input_set.tallies, merge=False)
self._input_set.tallies.export_to_xml()
# Add tallies
self.mgxs_lib.add_to_tallies_file(self._model.tallies, merge=False)
def _get_results(self, hash_output=True):
"""Digest info in the statepoint and return as a string."""
@ -65,5 +59,6 @@ class MGXSTestHarness(PyAPITestHarness):
if __name__ == '__main__':
harness = MGXSTestHarness('statepoint.10.h5', True)
model = pwr_pin_cell()
harness = MGXSTestHarness('statepoint.10.h5', model)
harness.main()

3
tests/test_multipole/inputs_true.dat
View file

@ -43,9 +43,6 @@
<parameters>-1 -1 -1 1 1 1</parameters>
</space>
</source>
<output>
<summary>true</summary>
</output>
<temperature_multipole>True</temperature_multipole>
<temperature_tolerance>1000</temperature_tolerance>
</settings>

47
tests/test_multipole/test_multipole.py
View file

@ -4,8 +4,6 @@ import sys
sys.path.insert(0, os.pardir)
from testing_harness import TestHarness, PyAPITestHarness
import openmc
from openmc.stats import Box
from openmc.source import Source
class MultipoleTestHarness(PyAPITestHarness):
def _build_inputs(self):
@ -66,8 +64,8 @@ class MultipoleTestHarness(PyAPITestHarness):
sets_file.batches = 5
sets_file.inactive = 0
sets_file.particles = 1000
sets_file.source = Source(space=Box([-1, -1, -1], [1, 1, 1]))
sets_file.output = {'summary': True}
sets_file.source = openmc.Source(space=openmc.stats.Box(
[-1, -1, -1], [1, 1, 1]))
sets_file.temperature = {'tolerance': 1000, 'multipole': True}
sets_file.export_to_xml()
@ -75,27 +73,24 @@ class MultipoleTestHarness(PyAPITestHarness):
# Plots
####################
plots_file = openmc.Plots()
plot1 = openmc.Plot(plot_id=1)
plot1.basis = 'xy'
plot1.color_by = 'cell'
plot1.filename = 'cellplot'
plot1.origin = (0, 0, 0)
plot1.width = (7, 7)
plot1.pixels = (400, 400)
plot = openmc.Plot(plot_id=1)
plot.basis = 'xy'
plot.color_by = 'cell'
plot.filename = 'cellplot'
plot.origin = (0, 0, 0)
plot.width = (7, 7)
plot.pixels = (400, 400)
plots_file.append(plot)
plot2 = openmc.Plot(plot_id=2)
plot2.basis = 'xy'
plot2.color_by = 'material'
plot2.filename = 'matplot'
plot2.origin = (0, 0, 0)
plot2.width = (7, 7)
plot2.pixels = (400, 400)
plot = openmc.Plot(plot_id=2)
plot.basis = 'xy'
plot.color_by = 'material'
plot.filename = 'matplot'
plot.origin = (0, 0, 0)
plot.width = (7, 7)
plot.pixels = (400, 400)
plots_file.append(plot)
plots_file.export_to_xml()
plots = openmc.Plots([plot1, plot2])
plots.export_to_xml()
def execute_test(self):
if not 'OPENMC_MULTIPOLE_LIBRARY' in os.environ:
@ -110,12 +105,6 @@ class MultipoleTestHarness(PyAPITestHarness):
outstr += str(su.geometry.get_all_cells()[11])
return outstr
def _cleanup(self):
f = os.path.join(os.getcwd(), 'plots.xml')
if os.path.exists(f):
os.remove(f)
super(MultipoleTestHarness, self)._cleanup()
if __name__ == '__main__':
harness = MultipoleTestHarness('statepoint.5.h5')

2
tests/test_output/test_output.py
View file

@ -29,5 +29,5 @@ class OutputTestHarness(TestHarness):
if __name__ == '__main__':
harness = OutputTestHarness('statepoint.10.*')
harness = OutputTestHarness('statepoint.10.h5')
harness.main()

2
tests/test_plot/test_plot.py
View file

@ -15,7 +15,7 @@ import openmc
class PlotTestHarness(TestHarness):
"""Specialized TestHarness for running OpenMC plotting tests."""
def __init__(self, plot_names):
super(PlotTestHarness, self).__init__(None, False)
super(PlotTestHarness, self).__init__(None)
self._plot_names = plot_names
def _run_openmc(self):

2
tests/test_ptables_off/test_ptables_off.py
View file

@ -7,5 +7,5 @@ from testing_harness import TestHarness
if __name__ == '__main__':
harness = TestHarness('statepoint.10.*')
harness = TestHarness('statepoint.10.h5')
harness.main()

2
tests/test_quadric_surfaces/test_quadric_surfaces.py
View file

@ -7,5 +7,5 @@ from testing_harness import TestHarness
if __name__ == '__main__':
harness = TestHarness('statepoint.10.*')
harness = TestHarness('statepoint.10.h5')
harness.main()

2
tests/test_reflective_plane/test_reflective_plane.py
View file

@ -7,5 +7,5 @@ from testing_harness import TestHarness
if __name__ == '__main__':
harness = TestHarness('statepoint.10.*')
harness = TestHarness('statepoint.10.h5')
harness.main()

2
tests/test_rotation/test_rotation.py
View file

@ -7,5 +7,5 @@ from testing_harness import TestHarness
if __name__ == '__main__':
harness = TestHarness('statepoint.10.*')
harness = TestHarness('statepoint.10.h5')
harness.main()

2
tests/test_salphabeta/test_salphabeta.py
View file

@ -7,5 +7,5 @@ from testing_harness import TestHarness
if __name__ == '__main__':
harness = TestHarness('statepoint.10.*')
harness = TestHarness('statepoint.10.h5')
harness.main()

2
tests/test_score_current/test_score_current.py
View file

@ -7,5 +7,5 @@ from testing_harness import HashedTestHarness
if __name__ == '__main__':
harness = HashedTestHarness('statepoint.10.*', True)
harness = HashedTestHarness('statepoint.10.h5')
harness.main()

2
tests/test_seed/test_seed.py
View file

@ -7,5 +7,5 @@ from testing_harness import TestHarness
if __name__ == '__main__':
harness = TestHarness('statepoint.10.*')
harness = TestHarness('statepoint.10.h5')
harness.main()

2
tests/test_source_file/test_source_file.py
View file

@ -98,5 +98,5 @@ class SourceFileTestHarness(TestHarness):
if __name__ == '__main__':
harness = SourceFileTestHarness('statepoint.10.*')
harness = SourceFileTestHarness('statepoint.10.h5')
harness.main()

19
tests/test_sourcepoint_batch/test_sourcepoint_batch.py
View file

@ -5,7 +5,7 @@ import os
import sys
sys.path.insert(0, os.pardir)
from testing_harness import TestHarness
from openmc.statepoint import StatePoint
from openmc import StatePoint
class SourcepointTestHarness(TestHarness):
@ -18,21 +18,20 @@ class SourcepointTestHarness(TestHarness):
def _get_results(self):
"""Digest info in the statepoint and return as a string."""
# Read the statepoint file.
statepoint = glob.glob(os.path.join(os.getcwd(), self._sp_name))[0]
sp = StatePoint(statepoint)
# Get the eigenvalue information.
outstr = TestHarness._get_results(self)
# Add the source information.
xyz = sp.source[0]['xyz']
outstr += ' '.join(['{0:12.6E}'.format(x) for x in xyz])
outstr += "\n"
# Read the statepoint file.
statepoint = glob.glob(os.path.join(os.getcwd(), self._sp_name))[0]
with StatePoint(statepoint) as sp:
# Add the source information.
xyz = sp.source[0]['xyz']
outstr += ' '.join(['{0:12.6E}'.format(x) for x in xyz])
outstr += "\n"
return outstr
if __name__ == '__main__':
harness = SourcepointTestHarness('statepoint.08.*')
harness = SourcepointTestHarness('statepoint.08.h5')
harness.main()

2
tests/test_sourcepoint_latest/test_sourcepoint_latest.py
View file

@ -18,5 +18,5 @@ class SourcepointTestHarness(TestHarness):
if __name__ == '__main__':
harness = TestHarness('statepoint.10.*')
harness = TestHarness('statepoint.10.h5')
harness.main()

2
tests/test_sourcepoint_restart/test_sourcepoint_restart.py
View file

@ -7,5 +7,5 @@ from testing_harness import TestHarness
if __name__ == '__main__':
harness = TestHarness('statepoint.10.*', True)
harness = TestHarness('statepoint.10.h5')
harness.main()

6
tests/test_statepoint_batch/test_statepoint_batch.py
View file

@ -1,4 +1,4 @@
#!/usr/bin/env python
#!/usr/bin/env python
import os
import sys
@ -8,11 +8,11 @@ from testing_harness import TestHarness
class StatepointTestHarness(TestHarness):
def __init__(self):
super(StatepointTestHarness, self).__init__(None, False)
super(StatepointTestHarness, self).__init__(None)
def _test_output_created(self):
"""Make sure statepoint files have been created."""
sps = ('statepoint.03.*', 'statepoint.06.*', 'statepoint.09.*')
sps = ('statepoint.03.h5', 'statepoint.06.h5', 'statepoint.09.h5')
for sp in sps:
self._sp_name = sp
TestHarness._test_output_created(self)

7
tests/test_statepoint_restart/test_statepoint_restart.py
View file

@ -9,9 +9,8 @@ import openmc
class StatepointRestartTestHarness(TestHarness):
def __init__(self, final_sp, restart_sp, tallies_present=False):
super(StatepointRestartTestHarness, self).__init__(final_sp,
tallies_present)
def __init__(self, final_sp, restart_sp):
super(StatepointRestartTestHarness, self).__init__(final_sp)
self._restart_sp = restart_sp
def execute_test(self):
@ -63,5 +62,5 @@ class StatepointRestartTestHarness(TestHarness):
if __name__ == '__main__':
harness = StatepointRestartTestHarness('statepoint.10.h5',
'statepoint.07.h5', True)
'statepoint.07.h5')
harness.main()

2
tests/test_statepoint_sourcesep/test_statepoint_sourcesep.py
View file

@ -26,5 +26,5 @@ class SourcepointTestHarness(TestHarness):
if __name__ == '__main__':
harness = SourcepointTestHarness('statepoint.10.*')
harness = SourcepointTestHarness('statepoint.10.h5')
harness.main()

2
tests/test_survival_biasing/test_survival_biasing.py
View file

@ -7,5 +7,5 @@ from testing_harness import TestHarness
if __name__ == '__main__':
harness = TestHarness('statepoint.10.*', True)
harness = TestHarness('statepoint.10.h5')
harness.main()

6
tests/test_tallies/inputs_true.dat
View file

@ -148,7 +148,7 @@
</geometry>
<?xml version='1.0' encoding='utf-8'?>
<materials>
<material id="1" name="Fuel">
<material id="1" name="UOX fuel">
<density units="g/cm3" value="10.062" />
<nuclide ao="4.9476e-06" name="U234" />
<nuclide ao="0.00048218" name="U235" />
@ -197,7 +197,7 @@
<nuclide name="Zr96" wo="0.0216912612" />
<sab name="c_H_in_H2O" />
</material>
<material id="2" name="Cladding">
<material id="2" name="Zircaloy">
<density units="g/cm3" value="5.77" />
<nuclide ao="0.5145" name="Zr90" />
<nuclide ao="0.1122" name="Zr91" />
@ -249,7 +249,7 @@
<nuclide name="Cr52" wo="0.145407678031" />
<sab name="c_H_in_H2O" />
</material>
<material id="7" name="Upper radial reflector /Top plate region">
<material id="7" name="Upper radial reflector / Top plate region">
<density units="g/cm3" value="4.28" />
<nuclide name="H1" wo="0.0086117" />
<nuclide name="O16" wo="0.0683369" />

330
tests/test_tallies/test_tallies.py
View file

@ -4,180 +4,168 @@ import os
import sys
sys.path.insert(0, os.pardir)
from testing_harness import PyAPITestHarness
from testing_harness import HashedPyAPITestHarness
from openmc.filter import *
from openmc import Mesh, Tally, Tallies
from openmc.source import Source
from openmc.stats import Box
class TalliesTestHarness(PyAPITestHarness):
def _build_inputs(self):
# Build default materials/geometry
self._input_set.build_default_materials_and_geometry()
# Set settings explicitly
self._input_set.settings.batches = 5
self._input_set.settings.inactive = 0
self._input_set.settings.particles = 400
self._input_set.settings.source = Source(space=Box(
[-160, -160, -183], [160, 160, 183]))
azimuthal_bins = (-3.14159, -1.8850, -0.6283, 0.6283, 1.8850, 3.14159)
azimuthal_filter = AzimuthalFilter(azimuthal_bins)
azimuthal_tally1 = Tally()
azimuthal_tally1.filters = [azimuthal_filter]
azimuthal_tally1.scores = ['flux']
azimuthal_tally1.estimator = 'tracklength'
azimuthal_tally2 = Tally()
azimuthal_tally2.filters = [azimuthal_filter]
azimuthal_tally2.scores = ['flux']
azimuthal_tally2.estimator = 'analog'
mesh_2x2 = Mesh(mesh_id=1)
mesh_2x2.lower_left = [-182.07, -182.07]
mesh_2x2.upper_right = [182.07, 182.07]
mesh_2x2.dimension = [2, 2]
mesh_filter = MeshFilter(mesh_2x2)
azimuthal_tally3 = Tally()
azimuthal_tally3.filters = [azimuthal_filter, mesh_filter]
azimuthal_tally3.scores = ['flux']
azimuthal_tally3.estimator = 'tracklength'
cellborn_tally = Tally()
cellborn_tally.filters = [CellbornFilter((10, 21, 22, 23))]
cellborn_tally.scores = ['total']
dg_tally = Tally()
dg_tally.filters = [DelayedGroupFilter((1, 2, 3, 4, 5, 6))]
dg_tally.scores = ['delayed-nu-fission']
four_groups = (0.0, 0.253, 1.0e3, 1.0e6, 20.0e6)
energy_filter = EnergyFilter(four_groups)
energy_tally = Tally()
energy_tally.filters = [energy_filter]
energy_tally.scores = ['total']
energyout_filter = EnergyoutFilter(four_groups)
energyout_tally = Tally()
energyout_tally.filters = [energyout_filter]
energyout_tally.scores = ['scatter']
transfer_tally = Tally()
transfer_tally.filters = [energy_filter, energyout_filter]
transfer_tally.scores = ['scatter', 'nu-fission']
material_tally = Tally()
material_tally.filters = [MaterialFilter((1, 2, 3, 4))]
material_tally.scores = ['total']
mu_bins = (-1.0, -0.5, 0.0, 0.5, 1.0)
mu_filter = MuFilter(mu_bins)
mu_tally1 = Tally()
mu_tally1.filters = [mu_filter]
mu_tally1.scores = ['scatter', 'nu-scatter']
mu_tally2 = Tally()
mu_tally2.filters = [mu_filter, mesh_filter]
mu_tally2.scores = ['scatter', 'nu-scatter']
polar_bins = (0.0, 0.6283, 1.2566, 1.8850, 2.5132, 3.14159)
polar_filter = PolarFilter(polar_bins)
polar_tally1 = Tally()
polar_tally1.filters = [polar_filter]
polar_tally1.scores = ['flux']
polar_tally1.estimator = 'tracklength'
polar_tally2 = Tally()
polar_tally2.filters = [polar_filter]
polar_tally2.scores = ['flux']
polar_tally2.estimator = 'analog'
polar_tally3 = Tally()
polar_tally3.filters = [polar_filter, mesh_filter]
polar_tally3.scores = ['flux']
polar_tally3.estimator = 'tracklength'
universe_tally = Tally()
universe_tally.filters = [UniverseFilter((1, 2, 3, 4, 6, 8))]
universe_tally.scores = ['total']
cell_filter = CellFilter((10, 21, 22, 23, 60))
score_tallies = [Tally(), Tally(), Tally()]
for t in score_tallies:
t.filters = [cell_filter]
t.scores = ['absorption', 'delayed-nu-fission', 'events', 'fission',
'inverse-velocity', 'kappa-fission', '(n,2n)', '(n,n1)',
'(n,gamma)', 'nu-fission', 'scatter', 'elastic',
'total', 'prompt-nu-fission', 'fission-q-prompt',
'fission-q-recoverable']
score_tallies[0].estimator = 'tracklength'
score_tallies[1].estimator = 'analog'
score_tallies[2].estimator = 'collision'
cell_filter2 = CellFilter((21, 22, 23, 27, 28, 29, 60))
flux_tallies = [Tally() for i in range(4)]
for t in flux_tallies:
t.filters = [cell_filter2]
flux_tallies[0].scores = ['flux']
for t in flux_tallies[1:]:
t.scores = ['flux-y5']
flux_tallies[1].estimator = 'tracklength'
flux_tallies[2].estimator = 'analog'
flux_tallies[3].estimator = 'collision'
scatter_tally1 = Tally()
scatter_tally1.filters = [cell_filter]
scatter_tally1.scores = ['scatter', 'scatter-1', 'scatter-2', 'scatter-3',
'scatter-4', 'nu-scatter', 'nu-scatter-1',
'nu-scatter-2', 'nu-scatter-3', 'nu-scatter-4']
scatter_tally2 = Tally()
scatter_tally2.filters = [cell_filter]
scatter_tally2.scores = ['scatter-p4', 'scatter-y4', 'nu-scatter-p4',
'nu-scatter-y3']
total_tallies = [Tally() for i in range(4)]
for t in total_tallies:
t.filters = [cell_filter]
total_tallies[0].scores = ['total']
for t in total_tallies[1:]:
t.scores = ['total-y4']
t.nuclides = ['U235', 'total']
total_tallies[1].estimator = 'tracklength'
total_tallies[2].estimator = 'analog'
total_tallies[3].estimator = 'collision'
all_nuclide_tallies = [Tally() for i in range(4)]
for t in all_nuclide_tallies:
t.filters = [cell_filter]
t.estimator = 'tracklength'
t.nuclides = ['all']
t.scores = ['total']
all_nuclide_tallies[1].estimator = 'collision'
all_nuclide_tallies[2].filters = [mesh_filter]
all_nuclide_tallies[3].filters = [mesh_filter]
all_nuclide_tallies[3].nuclides = ['U235']
self._input_set.tallies = Tallies()
self._input_set.tallies += (
[azimuthal_tally1, azimuthal_tally2, azimuthal_tally3,
cellborn_tally, dg_tally, energy_tally, energyout_tally,
transfer_tally, material_tally, mu_tally1, mu_tally2,
polar_tally1, polar_tally2, polar_tally3, universe_tally])
self._input_set.tallies += score_tallies
self._input_set.tallies += flux_tallies
self._input_set.tallies += (scatter_tally1, scatter_tally2)
self._input_set.tallies += total_tallies
self._input_set.tallies += all_nuclide_tallies
self._input_set.export()
def _get_results(self):
return super(TalliesTestHarness, self)._get_results(hash_output=True)
if __name__ == '__main__':
harness = TalliesTestHarness('statepoint.5.h5', True)
harness = HashedPyAPITestHarness('statepoint.5.h5')
model = harness._model
# Set settings explicitly
model.settings.batches = 5
model.settings.inactive = 0
model.settings.particles = 400
model.settings.source = openmc.Source(space=openmc.stats.Box(
[-160, -160, -183], [160, 160, 183]))
azimuthal_bins = (-3.14159, -1.8850, -0.6283, 0.6283, 1.8850, 3.14159)
azimuthal_filter = AzimuthalFilter(azimuthal_bins)
azimuthal_tally1 = Tally()
azimuthal_tally1.filters = [azimuthal_filter]
azimuthal_tally1.scores = ['flux']
azimuthal_tally1.estimator = 'tracklength'
azimuthal_tally2 = Tally()
azimuthal_tally2.filters = [azimuthal_filter]
azimuthal_tally2.scores = ['flux']
azimuthal_tally2.estimator = 'analog'
mesh_2x2 = Mesh(mesh_id=1)
mesh_2x2.lower_left = [-182.07, -182.07]
mesh_2x2.upper_right = [182.07, 182.07]
mesh_2x2.dimension = [2, 2]
mesh_filter = MeshFilter(mesh_2x2)
azimuthal_tally3 = Tally()
azimuthal_tally3.filters = [azimuthal_filter, mesh_filter]
azimuthal_tally3.scores = ['flux']
azimuthal_tally3.estimator = 'tracklength'
cellborn_tally = Tally()
cellborn_tally.filters = [CellbornFilter((10, 21, 22, 23))]
cellborn_tally.scores = ['total']
dg_tally = Tally()
dg_tally.filters = [DelayedGroupFilter((1, 2, 3, 4, 5, 6))]
dg_tally.scores = ['delayed-nu-fission']
four_groups = (0.0, 0.253, 1.0e3, 1.0e6, 20.0e6)
energy_filter = EnergyFilter(four_groups)
energy_tally = Tally()
energy_tally.filters = [energy_filter]
energy_tally.scores = ['total']
energyout_filter = EnergyoutFilter(four_groups)
energyout_tally = Tally()
energyout_tally.filters = [energyout_filter]
energyout_tally.scores = ['scatter']
transfer_tally = Tally()
transfer_tally.filters = [energy_filter, energyout_filter]
transfer_tally.scores = ['scatter', 'nu-fission']
material_tally = Tally()
material_tally.filters = [MaterialFilter((1, 2, 3, 4))]
material_tally.scores = ['total']
mu_bins = (-1.0, -0.5, 0.0, 0.5, 1.0)
mu_filter = MuFilter(mu_bins)
mu_tally1 = Tally()
mu_tally1.filters = [mu_filter]
mu_tally1.scores = ['scatter', 'nu-scatter']
mu_tally2 = Tally()
mu_tally2.filters = [mu_filter, mesh_filter]
mu_tally2.scores = ['scatter', 'nu-scatter']
polar_bins = (0.0, 0.6283, 1.2566, 1.8850, 2.5132, 3.14159)
polar_filter = PolarFilter(polar_bins)
polar_tally1 = Tally()
polar_tally1.filters = [polar_filter]
polar_tally1.scores = ['flux']
polar_tally1.estimator = 'tracklength'
polar_tally2 = Tally()
polar_tally2.filters = [polar_filter]
polar_tally2.scores = ['flux']
polar_tally2.estimator = 'analog'
polar_tally3 = Tally()
polar_tally3.filters = [polar_filter, mesh_filter]
polar_tally3.scores = ['flux']
polar_tally3.estimator = 'tracklength'
universe_tally = Tally()
universe_tally.filters = [UniverseFilter((1, 2, 3, 4, 6, 8))]
universe_tally.scores = ['total']
cell_filter = CellFilter((10, 21, 22, 23, 60))
score_tallies = [Tally(), Tally(), Tally()]
for t in score_tallies:
t.filters = [cell_filter]
t.scores = ['absorption', 'delayed-nu-fission', 'events', 'fission',
'inverse-velocity', 'kappa-fission', '(n,2n)', '(n,n1)',
'(n,gamma)', 'nu-fission', 'scatter', 'elastic',
'total', 'prompt-nu-fission', 'fission-q-prompt',
'fission-q-recoverable']
score_tallies[0].estimator = 'tracklength'
score_tallies[1].estimator = 'analog'
score_tallies[2].estimator = 'collision'
cell_filter2 = CellFilter((21, 22, 23, 27, 28, 29, 60))
flux_tallies = [Tally() for i in range(4)]
for t in flux_tallies:
t.filters = [cell_filter2]
flux_tallies[0].scores = ['flux']
for t in flux_tallies[1:]:
t.scores = ['flux-y5']
flux_tallies[1].estimator = 'tracklength'
flux_tallies[2].estimator = 'analog'
flux_tallies[3].estimator = 'collision'
scatter_tally1 = Tally()
scatter_tally1.filters = [cell_filter]
scatter_tally1.scores = ['scatter', 'scatter-1', 'scatter-2', 'scatter-3',
'scatter-4', 'nu-scatter', 'nu-scatter-1',
'nu-scatter-2', 'nu-scatter-3', 'nu-scatter-4']
scatter_tally2 = Tally()
scatter_tally2.filters = [cell_filter]
scatter_tally2.scores = ['scatter-p4', 'scatter-y4', 'nu-scatter-p4',
'nu-scatter-y3']
total_tallies = [Tally() for i in range(4)]
for t in total_tallies:
t.filters = [cell_filter]
total_tallies[0].scores = ['total']
for t in total_tallies[1:]:
t.scores = ['total-y4']
t.nuclides = ['U235', 'total']
total_tallies[1].estimator = 'tracklength'
total_tallies[2].estimator = 'analog'
total_tallies[3].estimator = 'collision'
all_nuclide_tallies = [Tally() for i in range(4)]
for t in all_nuclide_tallies:
t.filters = [cell_filter]
t.estimator = 'tracklength'
t.nuclides = ['all']
t.scores = ['total']
all_nuclide_tallies[1].estimator = 'collision'
all_nuclide_tallies[2].filters = [mesh_filter]
all_nuclide_tallies[3].filters = [mesh_filter]
all_nuclide_tallies[3].nuclides = ['U235']
model.tallies += [
azimuthal_tally1, azimuthal_tally2, azimuthal_tally3,
cellborn_tally, dg_tally, energy_tally, energyout_tally,
transfer_tally, material_tally, mu_tally1, mu_tally2,
polar_tally1, polar_tally2, polar_tally3, universe_tally]
model.tallies += score_tallies
model.tallies += flux_tallies
model.tallies += (scatter_tally1, scatter_tally2)
model.tallies += total_tallies
model.tallies += all_nuclide_tallies
harness.main()

Some files were not shown because too many files have changed in this diff Show more