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Merge branch 'release-0.7.0'

Browse source
This commit is contained in:
Paul Romano 2015-08-19 09:45:20 +07:00
commit 31c1f86f44
546 changed files with 33026 additions and 228766 deletions
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34
.gitignore vendored
View file

@ -8,8 +8,12 @@
# Compiler python objects
*.pyc
# OpenMC executable
src/openmc
# Python distribution
dist/
openmc.egg-info/
# Inputs generated from Python API
examples/python/**/*.xml
# emacs backups
*~
@ -22,7 +26,7 @@ docs/build
docs/source/_images/*.pdf
# Source build
src/build
build
# build from src/utils/setup.py
src/utils/build
@ -30,14 +34,32 @@ src/utils/build
# xml-fortran reader
src/xml-fortran/xmlreader
# Modules built from XML templates
src/templates/*.f90
# Test results error file
results_error.dat
# Test build files
tests/build/
tests/ctestscript.run
# HDF5 files
*.h5
# Build files
src/CMakeCache.txt
src/CMakeFiles/
src/bin/
src/cmake_install.cmake
src/install_manifest.txt
# Data downloaded from NNDC
data/nndc
#Images
*.ppm
# PyCharm project configuration files
.idea
.idea/*
# IPython notebook checkpoints
.ipynb_checkpoints

54
.travis.yml Normal file
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@ -0,0 +1,54 @@
sudo: false
language: python
python:
- "2.7"
- "3.4"
addons:
apt:
packages:
- gfortran
- g++
cache:
directories:
- $HOME/mpich_install
- $HOME/hdf5_install
- $HOME/phdf5_install
before_install:
# ============== Handle Python third-party packages ==============
- if [[ "$TRAVIS_PYTHON_VERSION" == "2.7" ]]; then
wget https://repo.continuum.io/miniconda/Miniconda-latest-Linux-x86_64.sh -O miniconda.sh;
else
wget https://repo.continuum.io/miniconda/Miniconda3-latest-Linux-x86_64.sh -O miniconda.sh;
fi
- bash miniconda.sh -b -p $HOME/miniconda
- export PATH="$HOME/miniconda/bin:$PATH"
- hash -r
- conda config --set always_yes yes --set changeps1 no
- conda update -q conda
- conda info -a
- conda create -q -n test-environment python=$TRAVIS_PYTHON_VERSION numpy scipy h5py
- source activate test-environment
# Install GCC, MPICH, HDF5, PHDF5
- ./tests/travis_install.sh
- export FC=gfortran
- export MPI_DIR=$HOME/mpich_install
- export PHDF5_DIR=$HOME/phdf5_install
- export HDF5_DIR=$HOME/hdf5_install
install: true
before_script:
- cd data
- git clone --branch=master git://github.com/bhermanmit/nndc_xs nndc_xs
- cat nndc_xs/nndc.tar.gza* | tar xzvf -
- rm -rf nndc_xs
- export CROSS_SECTIONS=$PWD/nndc/cross_sections.xml
- cd ..
script:
- cd tests
- export OMP_NUM_THREADS=3
- ./travis.sh
- cd ..

164
src/CMakeLists.txt → CMakeLists.txt
View file

@ -21,23 +21,27 @@ endif()
option(openmp "Enable shared-memory parallelism with OpenMP" OFF)
option(profile "Compile with profiling flags" OFF)
option(petsc "Enable PETSC for use in CMFD acceleration" OFF)
option(debug "Compile with debug flags" OFF)
option(optimize "Turn on all compiler optimization flags" OFF)
option(verbose "Create verbose Makefiles" OFF)
option(coverage "Compile with flags" OFF)
option(coverage "Compile with coverage analysis flags" OFF)
option(mpif08 "Use Fortran 2008 MPI interface" OFF)
if (verbose)
set(CMAKE_VERBOSE_MAKEFILE on)
endif()
# Maximum number of nested coordinates levels
set(maxcoord 10 CACHE STRING "Maximum number of nested coordinate levels")
add_definitions(-DMAX_COORD=${maxcoord})
#===============================================================================
# MPI for distributed-memory parallelism / HDF5 for binary output
#===============================================================================
set(MPI_ENABLED FALSE)
set(HDF5_ENABLED FALSE)
if($ENV{FC} MATCHES "mpi.*")
if($ENV{FC} MATCHES "mpi[^/]*$")
message("-- Detected MPI wrapper: $ENV{FC}")
add_definitions(-DMPI)
set(MPI_ENABLED TRUE)
@ -52,11 +56,24 @@ elseif($ENV{FC} MATCHES "h5pfc$")
set(HDF5_ENABLED TRUE)
endif()
# Check for Fortran 2008 MPI interface
if(MPI_ENABLED AND mpif08)
message("-- Using Fortran 2008 MPI bindings")
add_definitions(-DMPIF08)
endif()
#===============================================================================
# Set compile/link flags based on which compiler is being used
#===============================================================================
if(CMAKE_Fortran_COMPILER_ID STREQUAL "GNU")
# Make sure version is sufficient
execute_process(COMMAND ${CMAKE_Fortran_COMPILER} -dumpversion
OUTPUT_VARIABLE GCC_VERSION)
if(GCC_VERSION VERSION_LESS 4.6)
message(FATAL_ERROR "gfortran version must be 4.6 or higher")
endif()
# GNU Fortran compiler options
set(f90flags "-cpp -std=f2008 -fbacktrace")
if(debug)
@ -116,7 +133,8 @@ elseif(CMAKE_Fortran_COMPILER_ID STREQUAL "PGI")
elseif(CMAKE_Fortran_COMPILER_ID STREQUAL "XL")
# IBM XL compiler options
set(f90flags "-WF,-DNO_F2008 -O2")
set(f90flags "-O2")
add_definitions(-DNO_F2008)
if(debug)
set(f90flags "-g -C -qflag=i:i -u")
set(ldflags "-g")
@ -143,78 +161,6 @@ elseif(CMAKE_Fortran_COMPILER_ID STREQUAL "Cray")
endif()
#===============================================================================
# PETSc for CMFD functionality
#===============================================================================
set (PETSC_ENABLED FALSE)
if(petsc)
set(PETSC_ENABLED TRUE)
find_package(PETSc REQUIRED HINTS $ENV{PETSC_DIR}/conf)
find_library(libpetsc petsc $ENV{PETSC_DIR}/lib)
# If libfblas wasn't found, search the PETSc lib directory
if(PETSC_FBLAS_LIB STREQUAL "PETSC_FBLAS_LIB-NOTFOUND")
find_library(PETSC_FBLAS_LIB fblas $ENV{PETSC_DIR}/lib)
list(REMOVE_ITEM PETSC_PACKAGE_LIBS PETSC_FBLAS_LIB-NOTFOUND)
list(INSERT PETSC_PACKAGE_LIBS 0 ${PETSC_FBLAS_LIB})
endif()
# If libflapack wasn't found, search the PETSc lib directory
if(PETSC_FLAPACK_LIB STREQUAL "PETSC_FLAPACK_LIB-NOTFOUND")
find_library(PETSC_FLAPACK_LIB flapack $ENV{PETSC_DIR}/lib)
list(REMOVE_ITEM PETSC_PACKAGE_LIBS PETSC_FLAPACK_LIB-NOTFOUND)
list(INSERT PETSC_PACKAGE_LIBS 0 ${PETSC_FLAPACK_LIB})
endif()
# If libdl wasn't found, search /usr/lib64
if(PETSC_DL_LIB STREQUAL "PETSC_DL_LIB-NOTFOUND")
find_library(PETSC_DL_LIB libdl.so /usr/lib64)
list(REMOVE_ITEM PETSC_PACKAGE_LIBS PETSC_DL_LIB-NOTFOUND)
list(INSERT PETSC_PACKAGE_LIBS 0 ${PETSC_DL_LIB})
endif()
# If libm wasn't found, search /usr/lib64
if(PETSC_M_LIB STREQUAL "PETSC_M_LIB-NOTFOUND")
find_library(PETSC_M_LIB libm.so /usr/lib64)
list(REMOVE_ITEM PETSC_PACKAGE_LIBS PETSC_M_LIB-NOTFOUND)
list(INSERT PETSC_PACKAGE_LIBS 0 ${PETSC_M_LIB})
endif()
# If libpthread wasn't found, search /usr/lib64
if(PETSC_PTHREAD_LIB STREQUAL "PETSC_PTHREAD_LIB-NOTFOUND")
find_library(PETSC_PTHREAD_LIB libpthread.so /usr/lib64)
list(REMOVE_ITEM PETSC_PACKAGE_LIBS PETSC_PTHREAD_LIB-NOTFOUND)
list(INSERT PETSC_PACKAGE_LIBS 0 ${PETSC_PTHREAD_LIB})
endif()
# If librt wasn't found, search /usr/lib64
if(PETSC_RT_LIB STREQUAL "PETSC_RT_LIB-NOTFOUND")
find_library(PETSC_RT_LIB librt.so /usr/lib64)
list(REMOVE_ITEM PETSC_PACKAGE_LIBS PETSC_RT_LIB-NOTFOUND)
list(INSERT PETSC_PACKAGE_LIBS 0 ${PETSC_RT_LIB})
endif()
# If libssl wasn't found, search /usr/lib64
if(PETSC_SSL_LIB STREQUAL "PETSC_SSL_LIB-NOTFOUND")
find_library(PETSC_SSL_LIB libssl.so /usr/lib64)
list(REMOVE_ITEM PETSC_PACKAGE_LIBS PETSC_SSL_LIB-NOTFOUND)
list(INSERT PETSC_PACKAGE_LIBS 0 ${PETSC_SSL_LIB})
endif()
# If libcrypto wasn't found, search /usr/lib64
if(PETSC_CRYPTO_LIB STREQUAL "PETSC_CRYPTO_LIB-NOTFOUND")
find_library(PETSC_CRYPTO_LIB libcrypto.so /usr/lib64)
list(REMOVE_ITEM PETSC_PACKAGE_LIBS PETSC_CRYPTO_LIB-NOTFOUND)
list(INSERT PETSC_PACKAGE_LIBS 0 ${PETSC_CRYPTO_LIB})
endif()
message("-- Using PETSC: ${libpetsc}")
add_definitions(-DPETSC)
include_directories($ENV{PETSC_DIR}/include)
set(libraries "${libpetsc};${PETSC_PACKAGE_LIBS};${libraries}")
endif()
#===============================================================================
# git SHA1 hash
#===============================================================================
@ -234,21 +180,29 @@ endif()
# Only initialize git submodules if it is not there. User is responsible
# for future updates of fox xml submodule.
if(NOT EXISTS ${CMAKE_CURRENT_SOURCE_DIR}/xml/fox/.git)
message("-- Initializing/Updating FoX XML submodule...")
execute_process(COMMAND git submodule init
WORKING_DIRECTORY ${CMAKE_CURRENT_SOURCE_DIR}/..)
execute_process(COMMAND git submodule update
WORKING_DIRECTORY ${CMAKE_CURRENT_SOURCE_DIR}/..)
if(NOT EXISTS ${CMAKE_CURRENT_SOURCE_DIR}/src/xml/fox/.git)
if(NOT EXISTS ${CMAKE_CURRENT_SOURCE_DIR}/.git)
message("-- Cloning FoX XML git repository...")
execute_process(COMMAND git clone https://github.com/mit-crpg/fox.git src/xml/fox
WORKING_DIRECTORY ${CMAKE_CURRENT_SOURCE_DIR})
execute_process(COMMAND git checkout bdc852f4f43d969fb1b179cba79295c1e095a455
WORKING_DIRECTORY ${CMAKE_CURRENT_SOURCE_DIR}/src/xml/fox)
else()
message("-- Initializing/Updating FoX XML submodule...")
execute_process(COMMAND git submodule init
WORKING_DIRECTORY ${CMAKE_CURRENT_SOURCE_DIR})
execute_process(COMMAND git submodule update
WORKING_DIRECTORY ${CMAKE_CURRENT_SOURCE_DIR})
endif()
endif()
add_subdirectory(xml/fox)
add_subdirectory(src/xml/fox)
#===============================================================================
# Build OpenMC executable
#===============================================================================
set(program "openmc")
file(GLOB source *.F90 xml/openmc_fox.F90)
file(GLOB source src/*.F90 src/xml/openmc_fox.F90)
add_executable(${program} ${source})
target_link_libraries(${program} ${libraries} fox_dom)
set_target_properties(${program} PROPERTIES
@ -260,24 +214,16 @@ set_target_properties(${program} PROPERTIES
#===============================================================================
install(TARGETS ${program} RUNTIME DESTINATION bin)
install(PROGRAMS utils/statepoint_cmp.py
DESTINATION bin
RENAME statepoint_cmp)
install(PROGRAMS utils/statepoint_histogram.py
DESTINATION bin
RENAME statepoint_histogram)
install(PROGRAMS utils/statepoint_meshplot.py
DESTINATION bin
RENAME statepoint_meshplot)
install(FILES ../man/man1/openmc.1 DESTINATION share/man/man1)
install(FILES ../LICENSE DESTINATION "share/doc/${program}/copyright")
install(DIRECTORY src/relaxng DESTINATION share/openmc)
install(FILES man/man1/openmc.1 DESTINATION share/man/man1)
install(FILES LICENSE DESTINATION "share/doc/${program}/copyright")
find_package(PythonInterp)
if(PYTHONINTERP_FOUND)
install(CODE "execute_process(
COMMAND ${PYTHON_EXECUTABLE} setup.py install
--prefix=${CMAKE_INSTALL_PREFIX}
WORKING_DIRECTORY ${CMAKE_CURRENT_SOURCE_DIR}/utils)")
WORKING_DIRECTORY ${CMAKE_CURRENT_SOURCE_DIR})")
endif()
#===============================================================================
@ -288,15 +234,7 @@ endif()
include(CTest)
# Get a list of all the tests to run
file(GLOB_RECURSE TESTS ${CMAKE_CURRENT_SOURCE_DIR}/../tests/test_*.py)
# Check to see if PETSC is compiled for CMFD tests
if (NOT ${PETSC_ENABLED})
file(GLOB_RECURSE CMFD_TESTS ${CMAKE_CURRENT_SOURCE_DIR}/../tests/test_cmfd_jfnk.py)
foreach(cmfd_test in ${CMFD_TESTS})
list(REMOVE_ITEM TESTS ${cmfd_test})
endforeach(cmfd_test)
endif(NOT ${PETSC_ENABLED})
file(GLOB_RECURSE TESTS ${CMAKE_CURRENT_SOURCE_DIR}/tests/test_*.py)
# Check for MEM_CHECK and COVERAGE variables
if (DEFINED ENV{MEM_CHECK})
@ -349,6 +287,22 @@ foreach(test ${TESTS})
WORKING_DIRECTORY ${TEST_PATH}
COMMAND $<TARGET_FILE:openmc> -p ${TEST_PATH})
elseif(${test} MATCHES "test_filter_distribcell")
# Add each case for distribcell tests
add_test(NAME ${TEST_NAME}_case-1
WORKING_DIRECTORY ${TEST_PATH}/case-1
COMMAND $<TARGET_FILE:openmc> ${TEST_PATH}/case-1)
add_test(NAME ${TEST_NAME}_case-2
WORKING_DIRECTORY ${TEST_PATH}/case-2
COMMAND $<TARGET_FILE:openmc> ${TEST_PATH}/case-2)
add_test(NAME ${TEST_NAME}_case-3
WORKING_DIRECTORY ${TEST_PATH}/case-3
COMMAND $<TARGET_FILE:openmc> ${TEST_PATH}/case-3)
add_test(NAME ${TEST_NAME}_case-4
WORKING_DIRECTORY ${TEST_PATH}/case-4
COMMAND $<TARGET_FILE:openmc> ${TEST_PATH}/case-4)
# If a restart test is encounted, need to run with -r and restart file(s)
elseif(${test} MATCHES "restart")

0
src/CTestConfig.cmake → CTestConfig.cmake
View file

0
src/Makefile → Makefile
View file

30
data/get_nndc_data.py
View file

@ -8,6 +8,12 @@ import sys
import tarfile
import glob
import hashlib
import argparse
parser = argparse.ArgumentParser()
parser.add_argument('-b', '--batch', action = 'store_true',
help = 'supresses standard in')
args = parser.parse_args()
try:
from urllib.request import urlopen
@ -15,8 +21,8 @@ except ImportError:
from urllib2 import urlopen
cwd = os.getcwd()
sys.path.append(os.path.join(cwd, '..', 'src', 'utils'))
from convert_binary import ascii_to_binary
sys.path.insert(0, os.path.join(cwd, '..'))
from openmc.ace import ascii_to_binary
baseUrl = 'http://www.nndc.bnl.gov/endf/b7.1/aceFiles/'
files = ['ENDF-B-VII.1-neutron-293.6K.tar.gz',
@ -73,7 +79,7 @@ for f in files:
print('Verifying MD5 checksums...')
for f, checksum in zip(files, checksums):
downloadsum = hashlib.md5(open(f, 'r').read()).hexdigest()
downloadsum = hashlib.md5(open(f, 'rb').read()).hexdigest()
if downloadsum != checksum:
raise IOError("MD5 checksum for {} does not match. If this is your first "
"time receiving this message, please re-run the script. "
@ -118,10 +124,13 @@ shutil.copyfile('cross_sections_nndc.xml', 'nndc/cross_sections.xml')
# PROMPT USER TO DELETE .TAR.GZ FILES
# Ask user to delete
if sys.version_info[0] < 3:
response = raw_input('Delete *.tar.gz files? ([y]/n) ')
if not args.batch:
if sys.version_info[0] < 3:
response = raw_input('Delete *.tar.gz files? ([y]/n) ')
else:
response = input('Delete *.tar.gz files? ([y]/n) ')
else:
response = input('Delete *.tar.gz files? ([y]/n) ')
response = 'y'
# Delete files if requested
if not response or response.lower().startswith('y'):
@ -134,10 +143,13 @@ if not response or response.lower().startswith('y'):
# PROMPT USER TO CONVERT ASCII TO BINARY
# Ask user to convert
if sys.version_info[0] < 3:
response = raw_input('Convert ACE files to binary? ([y]/n) ')
if not args.batch:
if sys.version_info[0] < 3:
response = raw_input('Convert ACE files to binary? ([y]/n) ')
else:
response = input('Convert ACE files to binary? ([y]/n) ')
else:
response = input('Convert ACE files to binary? ([y]/n) ')
response = 'y'
# Convert files if requested
if not response or response.lower().startswith('y'):

780
docs/source/_images/hex_lat.svg Normal file
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@ -0,0 +1,780 @@
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<!DOCTYPE svg PUBLIC "-//W3C//DTD SVG 1.1//EN"
"http://www.w3.org/Graphics/SVG/1.1/DTD/svg11.dtd">
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<style type="text/css">
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25
docs/source/conf.py
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@ -17,13 +17,19 @@ import sys, os
# add these directories to sys.path here. If the directory is relative to the
# documentation root, use os.path.abspath to make it absolute, like shown here.
sys.path.insert(0, os.path.abspath('../sphinxext'))
sys.path.insert(0, os.path.abspath('../..'))
# -- General configuration -----------------------------------------------------
# Add any Sphinx extension module names here, as strings. They can be extensions
# coming with Sphinx (named 'sphinx.ext.*') or your custom ones.
extensions = ['sphinx.ext.pngmath', 'sphinxcontrib.tikz']
extensions = ['sphinx.ext.autodoc',
'sphinx.ext.napoleon',
'sphinx.ext.pngmath',
'sphinxcontrib.tikz',
'sphinx_numfig',
'notebook_sphinxext']
# Add any paths that contain templates here, relative to this directory.
templates_path = ['_templates']
@ -39,16 +45,16 @@ master_doc = 'index'
# General information about the project.
project = u'OpenMC'
copyright = u'2011-2014, Massachusetts Institute of Technology'
copyright = u'2011-2015, Massachusetts Institute of Technology'
# The version info for the project you're documenting, acts as replacement for
# |version| and |release|, also used in various other places throughout the
# built documents.
#
# The short X.Y version.
version = "0.6"
version = "0.7"
# The full version, including alpha/beta/rc tags.
release = "0.6.2"
release = "0.7.0"
# The language for content autogenerated by Sphinx. Refer to documentation
# for a list of supported languages.
@ -189,12 +195,19 @@ latex_documents = [
]
latex_elements = {
'preamble': '''
'preamble': r"""
\usepackage{enumitem}
\usepackage{amsfonts}
\usepackage{amsmath}
\setlistdepth{9}
\usepackage{tikz}
\usetikzlibrary{shapes,snakes,shadows,arrows,calc,decorations.markings,patterns,fit,matrix,spy}
'''
\usepackage{fixltx2e}
\hypersetup{bookmarksdepth=3}
\setcounter{tocdepth}{2}
\numberwithin{equation}{section}
""",
'printindex': r""
}
# The name of an image file (relative to this directory) to place at the top of

3
docs/source/developers.rst
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@ -11,7 +11,8 @@ Active development of the OpenMC Monte Carlo code is currently led by:
* `Nick Horelik <https://github.com/nhorelik>`_
* `Adam Nelson <https://github.com/nelsonag>`_
* `Jon Walsh <https://github.com/walshjon>`_
* `Sterling Harper <https://github.com/walshjon>`_
* `Sterling Harper <https://github.com/smharper>`_
* `Will Boyd <https://github.com/wbinventor>`_
* `Benoit Forget <http://web.mit.edu/nse/people/faculty/forget.html>`_
* `Kord Smith <http://web.mit.edu/nse/people/faculty/smith.html>`_
* `Andrew Siegel <http://www.mcs.anl.gov/about/people_detail.php?id=404>`_

62
docs/source/devguide/docbuild.rst Normal file
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@ -0,0 +1,62 @@
.. _devguide_docbuild:
=============================
Building Sphinx Documentation
=============================
In order to build the documentation in the ``docs`` directory, you will need to
have the Sphinx_ third-party Python package. The easiest way to install Sphinx
is via pip:
.. code-block:: sh
sudo pip install sphinx
Additionally, you will also need two Sphinx extensions for TikZ support and
numbering figures. The sphinxcontrib-tikz_ package should be installed directly
from the git repository as such:
.. code-block:: sh
sudo pip install https://bitbucket.org/philexander/tikz/get/HEAD.tar.gz
The Numfig_ package can be installed directly with pip:
.. code-block:: sh
sudo pip install sphinx-numfig
-----------------------------------
Building Documentation as a Webpage
-----------------------------------
To build the documentation as a webpage (what appears at
http://mit-crpg.github.io/openmc), simply go to the ``docs`` directory and run:
.. code-block:: sh
make html
-------------------------------
Building Documentation as a PDF
-------------------------------
To build PDF documentation, you will need to have a LaTeX distribution installed
on your computer as well as Inkscape_, which is used to convert .svg files to
.pdf files. Inkscape can be installed in a Debian-derivative with:
.. code-block:: sh
sudo apt-get install inkscape
One the pre-requisites are installed, simply go to the ``docs`` directory and
run:
.. code-block:: sh
make latexpdf
.. _Sphinx: http://sphinx-doc.org
.. _sphinxcontrib-tikz: https://bitbucket.org/philexander/tikz
.. _Numfig: https://pypi.python.org/pypi/sphinx_numfig
.. _Inkscape: https://inkscape.org

1
docs/source/devguide/index.rst
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@ -18,3 +18,4 @@ as debugging.
xml-parsing
statepoint
voxel
docbuild

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48
docs/source/devguide/styleguide.rst
View file

@ -8,7 +8,10 @@ In order to keep the OpenMC code base consistent in style, this guide specifies
a number of rules which should be adhered to when modified existing code or
adding new code in OpenMC.
-------------
-------
Fortran
-------
General Rules
-------------
@ -35,7 +38,6 @@ Don't use ``print *`` or ``write(*,*)``. If writing to a file, use a specific
unit. Writing to standard output or standard error should be handled by the
``write_message`` subroutine or functionality in the error module.
----------
Procedures
----------
@ -47,7 +49,6 @@ intent(in), intent(out), or intent(inout).
Include a comment describing what each argument to a procedure is.
---------
Variables
---------
@ -91,7 +92,7 @@ allocation instead. Use allocatable variables instead of pointer variables when
possible.
Shared/Module Variables
-----------------------
+++++++++++++++++++++++
Always put shared variables in modules. Access module variables through a
``use`` statement. Always use the ``only`` specifier on the ``use`` statement
@ -99,14 +100,12 @@ except for variables from the global, constants, and various header modules.
Never use ``equivalence`` statements, ``common`` blocks, or ``data`` statements.
-------------------------
Derived Types and Classes
-------------------------
Derived types and classes should have CamelCase names with words not separated
by underscores or hyphens.
-----------
Indentation
-----------
@ -126,14 +125,21 @@ program, subroutine, function, if, associate, etc. Emacs users should set the
variables f90-if-indent, f90-do-indent, f90-continuation-indent,
f90-type-indent, f90-associate-indent, and f90-program indent to 2.
Continuation lines should be indented by an extra 5 spaces. This is the default
value of f90-continuation-indent in Emacs.
Continuation lines should be indented by at least 5 spaces. They may be indented
more in order to make the content match the context. For example, either of
these are valid continuation indentations:
.. code-block:: fortran
local_xyz(1) = xyz(1) - (this % lower_left(1) + &
(i_xyz(1) - HALF)*this % pitch(1))
call which_data(scatt_type, get_scatt, get_nuscatt, get_chi_t, get_chi_p, &
get_chi_d, scatt_order)
-------------------------
Whitespace in Expressions
-------------------------
Use a single space between arguments to procedures.
Use a single space between arguments to procedures.
Avoid extraneous whitespace in the following situations:
@ -146,3 +152,25 @@ Avoid extraneous whitespace in the following situations:
Yes: if (variable == 2) then
No: if ( variable==2 ) then
Do not leave trailing whitespace at the end of a line.
------
Python
------
Style for Python code should follow PEP8_.
Docstrings for functions and methods should follow numpydoc_ style.
Python code should work with both Python 2.7+ and Python 3.0+.
Use of third-party Python packages should be limited to numpy_, scipy_, and
h5py_. Use of other third-party packages must be implemented as optional
dependencies rather than required dependencies.
.. _PEP8: https://www.python.org/dev/peps/pep-0008/
.. _numpydoc: https://github.com/numpy/numpy/blob/master/doc/HOWTO_DOCUMENT.rst.txt
.. _numpy: http://www.numpy.org/
.. _scipy: http://www.scipy.org/
.. _h5py: http://www.h5py.org/

4
docs/source/devguide/voxel.rst
View file

@ -4,9 +4,7 @@
Voxel Plot Binary File Specifications
=====================================
----------
Revision 1
----------
The current revision of the voxel plot binary file is 1.
**integer(4) n_voxels_x**

20
docs/source/devguide/workflow.rst
View file

@ -7,7 +7,7 @@ Development Workflow
Anyone wishing to make contributions to OpenMC should be fully acquianted and
comfortable working with git_ and GitHub_. We assume here that you have git
installed on your system, have a GitHub account, and have setup SSH keys to be
able to create/push to repositories on GitHub.
able to create/push to repositories on GitHub.
Overview
--------
@ -71,7 +71,7 @@ features and bug fixes. The general steps for contributing are as follows:
git checkout -b newbranch develop
3. Make your changes on the new branch that you intend to have included in
*develop*. If you have made other changes that should not be merged back,
*develop*. If you have made other changes that should not be merged back,
ensure that those changes are made on a different branch.
4. Issue a pull request from GitHub and select the *develop* branch of
@ -98,7 +98,7 @@ OpenMC Test Suite
The purpose of this test suite is to ensure that OpenMC compiles using various
combinations of compiler flags and options, and that all user input options can
be used successfully without breaking the code. The test suite is comprised of
be used successfully without breaking the code. The test suite is comprised of
regression tests where different types of input files are configured and the
full OpenMC code is executed. Results from simulations are compared with
expected results. The test suite is comprised of many build configurations
@ -114,7 +114,7 @@ download these cross sections please do the following:
.. code-block:: sh
cd ../data
python get_nndc.py
python get_nndc_data.py
export CROSS_SECTIONS=<path_to_data_folder>/nndc/cross_sections.xml
The test suite can be run on an already existing build using:
@ -143,19 +143,15 @@ variables should be set if the default paths are incorrect:
* **MPI_DIR** - The path to the MPI directory.
* Default - */opt/mpich/3.1-gnu*
* Default - */opt/mpich/3.1.3-gnu*
* **HDF5_DIR** - The path to the HDF5 directory.
* Default - */opt/hdf5/1.8.12-gnu*
* Default - */opt/hdf5/1.8.14-gnu*
* **PHDF5_DIR** - The path to the parallel HDF5 directory.
* Default - */opt/phdf5/1.8.12-gnu*
* **PETSC_DIR** - The path to the PETSc directory.
* Default - */opt/petsc/3.4.4-gnu*
* Default - */opt/phdf5/1.8.14-gnu*
To run the full test suite, the following command can be executed in the
tests directory:
@ -192,7 +188,7 @@ Adding tests to test suite
To add a new test to the test suite, create a sub-directory in the tests
directory that conforms to the regular expression *test_*. To configure
a test you need to add the following files to your new test directory,
a test you need to add the following files to your new test directory,
*test_name* for example:
* OpenMC input XML files

11
docs/source/index.rst
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@ -18,18 +18,21 @@ free to send a message to the User's Group `mailing list`_.
.. _Massachusetts Institute of Technology: http://web.mit.edu
.. _mailing list: https://groups.google.com/forum/?fromgroups=#!forum/openmc-users
--------
Contents
--------
.. only:: html
--------
Contents
--------
.. toctree::
:maxdepth: 1
quickinstall
releasenotes/index
releasenotes
methods/index
usersguide/index
devguide/index
pythonapi/index
publications
license
developers

44
docs/source/methods/cmfd.rst
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@ -28,7 +28,7 @@ directions. An example of this is shown in the following expression:
:label: not1
\sum\limits_{u\in(x,y,z)}\left\langle\overline{J}^{u,g}_{l+1/2,m,n}
\Delta_m^v\Delta_n^w\right\rangle
\Delta_m^v\Delta_n^w\right\rangle
Here, :math:`u` takes on each direction one at a time. The parameter :math:`J`
is surface area-averaged over the transverse indices :math:`m` and :math:`n`
@ -142,7 +142,7 @@ defined from MC tallies as follows:
{\left\langle\overline{\overline\phi}_{l,m,n}^h
\Delta_l^u\Delta_m^v\Delta_n^w\right\rangle}
and
and
.. math::
:label: xs3
@ -303,13 +303,13 @@ interior cell,
\sum_{u\in
x,y,x}\frac{1}{\Delta_l^u}\left[\left(-\tilde{D}_{l-1/2,m,n}^{u,g} -
\hat{D}_{l-1/2,m,n}^{u,g}\right)\overline{\overline{\phi}}_{l-1,m,n}^g\right.
+ \left(\tilde{D}_{l-1/2,m,n}^{u,g} +
\\ + \left(\tilde{D}_{l-1/2,m,n}^{u,g} +
\tilde{D}_{l+1/2,m,n}^{u,g} - \hat{D}_{l-1/2,m,n}^{u,g} +
\hat{D}_{l+1/2,m,n}^{u,g}\right)\overline{\overline{\phi}}_{l,m,n}^g
\\ +
\\ +
\left. \left(-\tilde{D}_{l+1/2,m,n}^{u,g} +
\hat{D}_{l+1/2,m,n}^{u,g}\right)\overline{\overline{\phi}}_{l+1,m,n}^g
\right] +
\right] \\ +
\overline{\overline\Sigma}_{t_{l,m,n}}^g\overline{\overline{\phi}}_{l,m,n}^g
- \sum\limits_{h=1}^G\overline{\overline{\nu_s\Sigma}}^{h\rightarrow
g}_{s_{l,m,n}}\overline{\overline{\phi}}_{l,m,n}^h =
@ -318,7 +318,7 @@ interior cell,
It should be noted that before substitution, eq. :eq:`eq_neut_bal` was divided
by the volume of the cell, :math:`\Delta_l^u\Delta_m^v\Delta_n^w`. Equation
:eq:`eq_cmfd_sys` can be represented in operator form as
:eq:`eq_cmfd_sys` can be represented in operator form as
.. math::
:label: eq_CMFDopers
@ -349,7 +349,7 @@ and energy group. This is represented as
p_{l,m,n}^g =
\frac{\sum_{h=1}^{G}\overline{\overline{\nu_f\Sigma}}^{h\rightarrow
g}_{f_{l,m,n}}\overline{\overline{\phi}}_{l,m,n}^h\Delta_l^u\Delta_m^v
\Delta_n^w}{\sum_n\sum_m\sum_l\sum_{h=1}^{G}\overline{
\Delta_n^w}{\sum_n\sum_m\sum_l\sum_{h=1}^{G}\overline{
\overline{\nu_f\Sigma}}^{h\rightarrow
g}_{f_{l,m,n}}\overline{\overline{\phi}}_{l,m,n}^h\Delta_l^u\Delta_m^v
\Delta_n^w}.
@ -439,7 +439,7 @@ by the user. Once all diffusion parameters are calculated, CMFD matrices are
formed where energy groups are the inner most iteration index. In OpenMC,
compressed row storage sparse matrices are used due to the sparsity of CMFD
operators. An example of this sparsity is shown for the 3-D BEAVRS model in
figures :ref:`fig_loss` and :ref:`fig_prod` [BEAVRS]_. These matrices represent
figures :num:`fig-loss` and :num:`fig-prod` [BEAVRS]_. These matrices represent
an assembly radial mesh, 24 cell mesh in the axial direction and two energy
groups. The loss matrix is 99.92% sparse and the production matrix is 99.99%
sparse. Although the loss matrix looks like it is tridiagonal, it is really a
@ -473,19 +473,19 @@ no fission neutrons appear with energies in the thermal group.
| \ :math:`\left\langle\overline{J}^{u,g}_{l\pm 1/2,m,n}\Delta_m^v\Delta_n^w\right\rangle` | current | mesh, energy |
+--------------------------------------------------------------------------------------------+----------------+---------------------------+
.. _fig_loss:
.. _fig-loss:
.. figure:: ../_images/loss.png
:scale: 50
Sparsity of Neutron Loss Operator
Sparsity of Neutron Loss Operator
.. _fig_prod:
.. _fig-prod:
.. figure:: ../_images/prod.png
:scale: 50
Sparsity of Neutron Production Operator
Sparsity of Neutron Production Operator
To solve the eigenvalue problem with these matrices, different source iteration
and linear solvers can be used. The most common source iteration solver used is
@ -511,7 +511,7 @@ implemented to obtain eigenvalue and multigroup fluxes as described in [Gill]_
and [Knoll]_. This method is not the primary one used, but has gotten recent
attention due to its coupling advantages to other physics such as thermal
hydraulics. Once multigroup fluxes are obtained, a normalized fission source is
calculated in the code using eq. :eq:`eq_cmfd_psrc` directly.
calculated in the code using eq. :eq:`eq_cmfd_psrc` directly.
The next step in the process is to compute weight adjustment factors. These are
calculated by taking the ratio of the expected number of neutrons from the CMFD
@ -523,11 +523,11 @@ the current MC source, OpenMC sums the statistical
weights of neutrons from the source bank on a given spatial and energy mesh.
Once weight adjustment factors were calculated, each neutron's statistical
weight in the source bank was modified according to its location and energy.
Examples of CMFD simulations using OpenMC can be found in [Herman_Thesis]_.
Examples of CMFD simulations using OpenMC can be found in [HermanThesis]_.
----------
References
----------
.. only:: html
.. rubric:: References
.. [BEAVRS] Nick Horelik, Bryan Herman. *Benchmark for Evaluation And Verification of Reactor
Simulations*. Massachusetts Institute of Technology, http://crpg.mit.edu/pub/beavrs
@ -536,23 +536,23 @@ References
.. [Gill] Daniel F. Gill. *Newton-Krylov methods for the solution of the k-eigenvalue problem in
multigroup neutronics calculations*. Ph.D. thesis, Pennsylvania State University, 2010.
.. [Hebert] Alain Hebert. *Applied reactor physics*. Presses Internationales Polytechnique,
.. [Hebert] Alain Hebert. *Applied reactor physics*. Presses Internationales Polytechnique,
Montreal, 2009.
.. [Herman] Bryan R. Herman, Benoit Forget, Kord Smith, and Brian N. Aviles. Improved
diffusion coefficients generated from Monte Carlo codes. In *Proceedings of M&C
2013*, Sun Valley, ID, USA, May 5 - 9, 2013.
.. [Herman_Thesis] Bryan R. Herman. *Monte Carlo and Thermal Hydraulic Coupling using
Low-Order Nonlinear Diffusion Acceleration*. Sc.D. thesis,
Massachusetts Institute of Technology, 2014.
.. [HermanThesis] Bryan R. Herman. *Monte Carlo and Thermal Hydraulic Coupling using
Low-Order Nonlinear Diffusion Acceleration*. Sc.D. thesis,
Massachusetts Institute of Technology, 2014.
.. [Knoll] D.A. Knoll, H. Park, and C. Newman. *Acceleration of k-eigenvalue/criticality
calculations using the Jacobian-free Newton-Krylov method*. Nuclear Science and
Engineering, 167:133–140, 2011.
.. [Park] H. Park, D.A. Knoll, and C.K. Newman. *Nonlinear acceleration of transport
criticality problems*. Nuclear Science and Engineering, 172:52–65, 2012.
criticality problems*. Nuclear Science and Engineering, 172:52–65, 2012.
.. [Rhodes] Joel Rhodes and Malte Edenius. *CASMO-4 --- A Fuel Assembly Burnup Program.
User’s Manual*. Studsvik of America, ssp-09/443-u rev 0, proprietary edition, 2001.

6
docs/source/methods/cross_sections.rst
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@ -63,9 +63,9 @@ Other Methods
A good survey of other energy grid techniques, including unionized energy grids,
can be found in a paper by Leppanen_.
----------
References
----------
.. only:: html
.. rubric:: References
.. [Brown] Forrest B. Brown, "New Hash-based Energy Lookup Algorithm for Monte
Carlo codes," LA-UR-14-24530, Los Alamos National Laboratory (2014).

110
docs/source/methods/geometry.rst
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@ -42,9 +42,11 @@ One can confirm that any point inside this sphere will correspond to
In OpenMC, every surface defined by the user is assigned an integer to uniquely
identify it. We can then refer to either of the two half-spaces created by a
surface by a combination of the unique ID of the surface and a positive/negative
sign. The following illustration shows an example of an ellipse with unique ID 1
sign. Figure :num:`fig-halfspace` shows an example of an ellipse with unique ID 1
dividing space into two half-spaces.
.. _fig-halfspace:
.. figure:: ../_images/halfspace.*
:align: center
:figclass: align-center
@ -57,9 +59,11 @@ to be defined by intersections, unions, and differences or half-spaces, OpenMC
is currently limited to cells defined only as intersections of
half-spaces. Thus, the specification of the cell must include a list of
half-space references whose intersection defines the region. The region is then
assigned a material defined elsewhere. The following illustration shows an
assigned a material defined elsewhere. Figure :num:`fig-union` shows an
example of a cell defined as the intersection of an ellipse and two planes.
.. _fig-union:
.. figure:: ../_images/union.*
:align: center
:figclass: align-center
@ -385,6 +389,106 @@ is found that contains the specified point.
.. _cell-contains:
----------------------
Finding a Lattice Tile
----------------------
If a particle is inside a lattice, its position inside the lattice must be
determined before assigning it to a cell. Throughout this section, the
volumetric units of the lattice will be referred to as "tiles". Tiles are
identified by thier indices, and the process of discovering which tile contains
the particle is referred to as "indexing".
Rectilinear Lattice Indexing
----------------------------
Indices are assigned to tiles in a rectilinear lattice based on the tile's
position along the :math:`x`, :math:`y`, and :math:`z` axes. Figure
:num:`fig-rect-lat` maps the indices for a 2D lattice. The indices, (1, 1),
map to the lower-left tile. (5, 1) and (5, 5) map to the lower-right and
upper-right tiles, respectively.
.. _fig-rect-lat:
.. figure:: ../_images/rect_lat.*
:align: center
:figclass: align-center
:width: 400px
Rectilinear lattice tile indices.
In general, a lattice tile is specified by the three indices,
:math:`(i_x, i_y, i_z)`. If a particle's current coordinates are
:math:`(x, y, z)` then the indices can be determined from these formulas:
.. math::
:label: rect_indexing
i_x = \left \lceil \frac{x - x_0}{p_0} \right \rceil
i_y = \left \lceil \frac{y - y_0}{p_1} \right \rceil
i_z = \left \lceil \frac{z - z_0}{p_2} \right \rceil
where :math:`(x_0, y_0, z_0)` are the coordinates to the lower-left-bottom
corner of the lattice, and :math:`p_0, p_1, p_2` are the pitches along the
:math:`x`, :math:`y`, and :math:`z` axes, respectively.
Hexagonal Lattice Indexing
--------------------------
A skewed coordinate system is used for indexing hexagonal lattice tiles.
Rather than a :math:`y`-axis, another axis is used that is rotated 30 degrees
counter-clockwise from the :math:`y`-axis. This axis is referred to as the
:math:`\alpha`-axis. Figure :num:`fig-hex-lat` shows how 2D hexagonal tiles
are mapped with the :math:`(x, \alpha)` basis. In this system, (0, 0) maps to
the center tile, (0, 2) to the top tile, and (2, -1) to the middle tile on the
right side.
.. _fig-hex-lat:
.. figure:: ../_images/hex_lat.*
:align: center
:figclass: align-center
:width: 400px
Hexagonal lattice tile indices.
Unfortunately, the indices cannot be determined with one simple formula as
before. Indexing requires a two-step process, a coarse step which determines a
set of four tiles that contains the particle and a fine step that determines
which of those four tiles actually contains the particle.
In the first step, indices are found using these formulas:
.. math::
:label: hex_indexing
\alpha = -\frac{x}{\sqrt{3}} + y
i_x^* = \left \lfloor \frac{x}{p_0 \sqrt{3} / 2} \right \rfloor
i_\alpha^* = \left \lfloor \frac{\alpha}{p_0} \right \rfloor
where :math:`p_0` is the lattice pitch (in the :math:`x`-:math:`y` plane). The
true index of the particle could be :math:`(i_x^*, i_\alpha^*)`,
:math:`(i_x^* + 1, i_\alpha^*)`, :math:`(i_x^*, i_\alpha^* + 1)`, or
:math:`(i_x^* + 1, i_\alpha^* + 1)`.
The second step selects the correct tile from that neighborhood of 4. OpenMC
does this by calculating the distance between the particle and the centers of
each of the 4 tiles, and then picking the closest tile. This works because
regular hexagonal tiles form a Voronoi tessellation which means that all of the
points within a tile are closest to the center of that same tile.
Indexing along the :math:`z`-axis uses the same method from rectilinear
lattices, i.e.
.. math::
:label: hex_indexing_z
i_z = \left \lceil \frac{z - z_0}{p_2} \right \rceil
----------------------------------------
Determining if a Coordinate is in a Cell
----------------------------------------

21
docs/source/methods/parallelization.rst
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@ -61,7 +61,7 @@ one master process that controls the scheduling of work and the remaining
processes wait to receive work from the master, process the work, and then send
their results to the master at the end of the simulation (or a source iteration
in the case of an eigenvalue calculation). This idea is illustrated in
:ref:`Figure 1 <figure-master-slave>`.
:ref:`figure-master-slave`.
.. _figure-master-slave:
@ -69,7 +69,7 @@ in the case of an eigenvalue calculation). This idea is illustrated in
:align: center
:figclass: align-center
**Figure 1**: Communication pattern in master-slave algorithm.
Communication pattern in master-slave algorithm.
Eigenvalue calculations are slightly more difficult to parallelize than fixed
source calculations since it is necessary to converge on the fission source
@ -117,8 +117,7 @@ Nearest Neighbors Algorithm
To reduce the amount of communication required in a fission bank synchronization
algorithm, it is desirable to move away from the typical master-slave algorithm
to an algorithm whereby the compute nodes communicate with one another only as
needed. This concept is illustrated in :ref:`Figure 2
<figure-nearest-neighbor>`.
needed. This concept is illustrated in :ref:`figure-nearest-neighbor`.
.. _figure-nearest-neighbor:
@ -126,7 +125,7 @@ needed. This concept is illustrated in :ref:`Figure 2
:align: center
:figclass: align-center
**Figure 2**: Communication pattern in nearest neighbor algorithm.
Communication pattern in nearest neighbor algorithm.
Since the source sites for each cycle are sampled from the fission sites banked
from the previous cycle, it is a common occurrence for a fission site to be
@ -196,8 +195,8 @@ and :math:`p_3` has 245. Note that the total number of sampled sites is 1000 as
needed. For each node to have the same number of source sites, :math:`p_0` needs
to send its right-most 10 sites to :math:`p_1`, and :math:`p_2` needs to send
its left-most 25 sites to :math:`p_1` and its right-most 5 sites to
:math:`p_3`. A schematic of this example is shown in :ref:`Figure 3
<figure-neighbor-example>`. The data local to each node is given a different
:math:`p_3`. A schematic of this example is shown in
:ref:`figure-neighbor-example`. The data local to each node is given a different
hatching, and the cross-hatched regions represent source sites that are
communicated between adjacent nodes.
@ -207,7 +206,7 @@ communicated between adjacent nodes.
:align: center
:figclass: align-center
**Figure 3**: Example of nearest neighbor algorithm.
Example of nearest neighbor algorithm.
.. _master-slave-cost:
@ -600,9 +599,9 @@ is actually independent of the number of nodes:
E \left [ \Lambda_{j_{\text{max}}} \right ] = \sqrt{ \frac{N\sigma^2}{2\pi
k^2}}.
----------
References
----------
.. only:: html
.. rubric:: References
.. [Troubetzkoy] E. Troubetzkoy, H. Steinberg, and M. Kalos, "Monte Carlo
Radiation Penetration Calculations on a Parallel Computer,"

6
docs/source/methods/physics.rst
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@ -1552,9 +1552,9 @@ default, the cutoff weight in OpenMC is :math:`w_c = 0.25` and the survival
weight is :math:`w_s = 1.0`. These parameters vary from one Monte Carlo code to
another.
----------
References
----------
.. only:: html
.. rubric:: References
.. [Doyas] Richard J. Doyas and Sterrett T. Perkins, "Interpolation of Tabular
Secondary Neutron and Photon Energy Distributions," *Nucl. Sci. Eng.*,

6
docs/source/methods/random_numbers.rst
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@ -62,9 +62,9 @@ Note that :eq:`lcg-skipahead` has the same general form as \eqref{eq:lcg}, so
the idea is to determine the new multiplicative and additive constants in
:math:`O(\log_2 N)` operations.
----------
References
----------
.. only:: html
.. rubric:: References
.. [LEcuyer] P. L’Ecuyer, "Tables of Linear Congruential Generators of
Different Sizes and Good Lattice Structures," *Math. Comput.*, **68**, 249

6
docs/source/methods/tallies.rst
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@ -475,9 +475,9 @@ normal distribution, we use an `unpublished rational approximation`_. After
using the rational approximation, one iteration of Newton's method is applied to
improve the estimate of the percentile.
----------
References
----------
.. only:: html
.. rubric:: References
.. [George] E. E. Olusegun George and Meenakshi Sivaram, "A modification of the
Fisher-Cornish approximation for the student t percentiles," Communication

160
docs/source/publications.rst
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@ -4,6 +4,104 @@
Publications
============
---------
Overviews
---------
- Paul K. Romano, Nicholas E. Horelik, Bryan R. Herman, Adam G. Nelson, Benoit
Forget, and Kord Smith, "OpenMC: A State-of-the-Art Monte Carlo Code for
Research and Development," *Ann. Nucl. Energy*, **82**, 90--97
(2015). `<http://dx.doi.org/10.1016/j.anucene.2014.07.048>`_
- Paul K. Romano, Bryan R. Herman, Nicholas E. Horelik, Benoit Forget, Kord
Smith, and Andrew R. Siegel, "Progress and Status of the OpenMC Monte Carlo
Code," *Proc. Int. Conf. Mathematics and Computational Methods Applied to
Nuclear Science and Engineering*, Sun Valley, Idaho, May 5--9 (2013).
- Paul K. Romano and Benoit Forget, "The OpenMC Monte Carlo Particle Transport
Code," *Ann. Nucl. Energy*, **51**, 274--281
(2013). `<http://dx.doi.org/10.1016/j.anucene.2012.06.040>`_
------------
Benchmarking
------------
- Daniel J. Kelly, Brian N. Aviles, Paul K. Romano, Bryan R. Herman,
Nicholas E. Horelik, and Benoit Forget, "Analysis of select BEAVRS PWR
benchmark cycle 1 results using MC21 and OpenMC," *Proc. PHYSOR*, Kyoto,
Japan, Sep. 28--Oct. 3 (2014).
- Bryan R. Herman, Benoit Forget, Kord Smith, Paul K. Romano, Thomas M. Sutton,
Daniel J. Kelly, III, and Brian N. Aviles, "Analysis of tally correlations in
large light water reactors," *Proc. PHYSOR*, Kyoto, Japan, Sep. 28--Oct. 3
(2014).
- Nicholas Horelik, Bryan Herman, Benoit Forget, and Kord Smith, "Benchmark for
Evaluation and Validation of Reactor Simulations,"
*Proc. Int. Conf. Mathematics and Computational Methods Applied to Nuclear
Science and Engineering*, Sun Valley, Idaho, May 5--9 (2013).
- Jonathan A. Walsh, Benoit Forget, and Kord S. Smith, "Validation of OpenMC
Reactor Physics Simulations with the B&W 1810 Series Benchmarks,"
*Trans. Am. Nucl. Soc.*, **109**, 1301--1304 (2013).
--------------------------
Coupling and Multi-physics
--------------------------
- Matt Ellis, Benoit Forget, Kord Smith, and Derek Gaston, "Preliminary coupling
of the Monte Carlo code OpenMC and the Multiphysics Object-Oriented Simulation
Environment (MOOSE) for analyzing Doppler feedback in Monte Carlo
simulations," *Proc. Joint Int. Conf. M&C+SNA+MC*, Nashville, Tennessee,
Apr. 19--23 (2015).
- Bryan R. Herman, Benoit Forget, and Kord Smith, "Progress toward Monte
Carlo-thermal hydraulic coupling using low-order nonlinear diffusion
acceleration methods." In press, *Ann. Nucl. Energy*,
(2014). `<http://dx.doi.org/10.1016/j.anucene/2014.10.029>`_
- Adam G. Nelson and William R. Martin, "Improved Convergence of Monte Carlo
Generated Multi-Group Scattering Moments," *Proc. Int. Conf. Mathematics and
Computational Methods Applied to Nuclear Science and Engineering*, Sun Valley,
Idaho, May 5--9 (2013).
- Bryan R. Herman, Benoit Forget, and Kord Smith, "Utilizing CMFD in OpenMC to
Estimate Dominance Ratio and Adjoint," *Trans. Am. Nucl. Soc.*, **109**,
1389-1392 (2013).
--------
Geometry
--------
- Derek Lax, William Boyd, Nicholas Horelik, Benoit Forget, and Kord Smith, "A
memory efficient algorithm for classifying unique regions in constructive
solid geometries," *Proc. PHYSOR*, Kyoto, Japan, Sep. 28--Oct. 3 (2014).
-------------
Miscellaneous
-------------
- Timothy P. Burke, Brian C. Kiedrowski, and William R. Martin, "Flux and
Reaction Rate Kernel Density Estimators in OpenMC," *Trans. Am. Nucl. Soc.*,
**109**, 683-686 (2013).
------------
Nuclear Data
------------
- Jonathan A. Walsh, Benoit Forget, Kord S. Smith, Brian C. Kiedrowski, and
Forrest B. Brown, "Direct, on-the-fly calculation of unresolved resonance
region cross sections in Monte Carlo simulations," *Proc. Joint
Int. Conf. M&C+SNA+MC*, Nashville, Tennessee, Apr. 19--23 (2015).
- Paul K. Romano and Timothy H. Trumbull, "Comparison of algorithms for Doppler
broadening pointwise tabulated cross sections," *Ann. Nucl. Energy*, **75**,
358--364 (2015). `<http://dx.doi.org/10.1016/j.anucene.2014.08.046>`_
- Tuomas Viitanen, Jaakko Leppanen, and Benoit Forget, "Target motion sampling
temperature treatment technique with track-length esimators in OpenMC --
Preliminary results," *Proc. PHYSOR*, Kyoto, Japan, Sep. 28--Oct. 3 (2014).
- Jonathan A. Walsh, Benoit Forget, and Kord S. Smith, "Accelerated sampling
of the free gas resonance elastic scattering kernel," *Ann. Nucl. Energy*,
**69**, 116--124 (2014). `<http://dx.doi.org/10.1016/j.anucene.2014.01.017>`_
@ -12,35 +110,42 @@ Publications
Carlo simulations using the multipole representation," *Ann. Nucl. Energy*,
**64**, 78--85 (2014). `<http://dx.doi.org/10.1016/j.anucene.2013.09.043>`_
-----------
Parallelism
-----------
- David Ozog, Allen D. Malony, and Andrew R. Siegel, "A performance analysis of
SIMD algorithms for Monte Carlo simulations of nuclear reactor cores,"
*Proc. IEEE Int. Parallel and Distributed Processing Symposium*, Hyderabad,
India, May 25--29 (2015).
- David Ozog, Allen D. Malony, and Andrew Siegel, "Full-core PWR transport
simulations on Xeon Phi clusters," *Proc. Joint Int. Conf. M&C+SNA+MC*,
Nashville, Tennessee, Apr. 19--23 (2015).
- Paul K. Romano, Andrew R. Siegel, and Ronald O. Rahaman, "Influence of the
memory subsystem on Monte Carlo code performance," *Proc. Joint
Int. Conf. M&C+SNA+MC*, Nashville, Tennessee, Apr. 19--23 (2015).
- Nicholas Horelik, Benoit Forget, Kord Smith, and Andrew Siegel, "Domain
decomposition and terabyte tallies with the OpenMC Monte Carlo neutron
transport code," *Proc. PHYSOR*, Kyoto Japan, Sep. 28--Oct. 3 (2014).
- Nicholas Horelik, Andrew Siegel, Benoit Forget, and Kord Smith, "Monte Carlo
domain decomposition for robust nuclear reactor analysis," *Parallel Comput.*,
**40**, 646--660 (2014). `<http://dx.doi.org/10.1016/j.parco.2014.10.001>`_
- Andrew Siegel, Kord Smith, Kyle Felker, Paul Romano, Benoit Forget, and Peter
Beckman, "Improved cache performance in Monte Carlo transport calculations
using energy banding," *Comput. Phys. Commun.*, **185** (4), 1195--1199
(2014). `<http://dx.doi.org/10.1016/j.cpc.2013.10.008>`_
- Jonathan A. Walsh, Benoit Forget, and Kord S. Smith, "Validation of OpenMC
Reactor Physics Simulations with the B&W 1810 Series Benchmarks,"
*Trans. Am. Nucl. Soc.*, **109**, 1301--1304 (2013).
- Bryan R. Herman, Benoit Forget, and Kord Smith, "Utilizing CMFD in OpenMC to
Estimate Dominance Ratio and Adjoint," *Trans. Am. Nucl. Soc.*, **109**,
1389-1392 (2013).
- Timothy P. Burke, Brian C. Kiedrowski, and William R. Martin, "Flux and
Reaction Rate Kernel Density Estimators in OpenMC," *Trans. Am. Nucl. Soc.*,
**109**, 683-686 (2013).
- Paul K. Romano, Benoit Forget, Kord Smith, and Andrew Siegel, "On the use of
tally servers in Monte Carlo simulations of light-water reactors,"
*Proc. Joint International Conference on Supercomputing in Nuclear
Applications and Monte Carlo*, Paris, France, Oct. 27--31
(2013). `<http://dx.doi.org/10.1051/snamc/201404301>`_
- Paul K. Romano, Nicholas E. Horelik, Bryan R. Herman, Adam G. Nelson, Benoit
Forget, and Kord Smith, "OpenMC: A State-of-the-Art Monte Carlo Code for
Research and Development," *Proc. Joint International Conference on
Supercomputing in Nuclear Applications and Monte Carlo*, Paris, France,
Oct. 27--31 (2013). `<http://dx.doi.org/10.1051/snamc/201406016>`_
- Kyle G. Felker, Andrew R. Siegel, Kord S. Smith, Paul K. Romano, and Benoit
Forget, "The energy band memory server algorithm for parallel Monte Carlo
calculations," *Proc. Joint International Conference on Supercomputing in
@ -62,25 +167,6 @@ Publications
servers," *J. Comput. Phys.*, **252**, 20--36
(2013). `<http://dx.doi.org/10.1016/j.jcp.2013.06.011>`_
- Paul K. Romano, Bryan R. Herman, Nicholas E. Horelik, Benoit Forget, Kord
Smith, and Andrew R. Siegel, "Progress and Status of the OpenMC Monte Carlo
Code," *Proc. Int. Conf. Mathematics and Computational Methods Applied to
Nuclear Science and Engineering*, Sun Valley, Idaho, May 5--9 (2013).
- Nicholas Horelik, Bryan Herman, Benoit Forget, and Kord Smith, "Benchmark for
Evaluation and Validation of Reactor Simulations,"
*Proc. Int. Conf. Mathematics and Computational Methods Applied to Nuclear
Science and Engineering*, Sun Valley, Idaho, May 5--9 (2013).
- Adam G. Nelson and William R. Martin, "Improved Convergence of Monte Carlo
Generated Multi-Group Scattering Moments," *Proc. Int. Conf. Mathematics and
Computational Methods Applied to Nuclear Science and Engineering*, Sun Valley,
Idaho, May 5--9 (2013).
- Paul K. Romano and Benoit Forget, "The OpenMC Monte Carlo Particle Transport
Code," *Ann. Nucl. Energy*, **51**, 274--281
(2013). `<http://dx.doi.org/10.1016/j.anucene.2012.06.040>`_
- Andrew R. Siegel, Kord Smith, Paul K. Romano, Benoit Forget, and Kyle Felker,
"The effect of load imbalances on the performance of Monte Carlo codes in LWR
analysis", *J. Comput. Phys.*, **235**, 901--911 (2013).

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@ -0,0 +1,8 @@
.. _pythonapi_ace:
==========
ACE Format
==========
.. automodule:: openmc.ace
:members:

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@ -0,0 +1,8 @@
.. _pythonapi_cmfd:
====
CMFD
====
.. automodule:: openmc.cmfd
:members:

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@ -0,0 +1,8 @@
.. _pythonapi_element:
=======
Element
=======
.. automodule:: openmc.element
:members:

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@ -0,0 +1,11 @@
=================
Pandas Dataframes
=================
.. only:: html
.. notebook:: pandas-dataframes.ipynb
.. only:: latex
IPython notebooks must be viewed in the online HTML documentation.

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@ -0,0 +1,11 @@
================
Tally Arithmetic
================
.. only:: html
.. notebook:: tally-arithmetic.ipynb
.. only:: latex
IPython notebooks must be viewed in the online HTML documentation.

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@ -0,0 +1,8 @@
.. _pythonapi_executor:
========
Executor
========
.. automodule:: openmc.executor
:members:

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@ -0,0 +1,8 @@
.. _pythonapi_filter:
======
Filter
======
.. automodule:: openmc.filter
:members:

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@ -0,0 +1,8 @@
.. _pythonapi_geometry:
========
Geometry
========
.. automodule:: openmc.geometry
:members:

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@ -0,0 +1,71 @@
.. _pythonapi:
==========
Python API
==========
OpenMC includes a rich Python API that enables programmatic pre- and
post-processing. The easiest way to begin using the API is to take a look at the
example Jupyter_ notebooks provided. However, this assumes that you are already
familiar with Python and common third-party packages such as NumPy_. If you have
never programmed in Python before, there are many good tutorials available
online. We recommend going through the modules from Codecademy_ and/or the
`Scipy lectures`_. The full API documentation serves to provide more information
on a given module or class.
**Handling nuclear data:**
.. toctree::
:maxdepth: 1
ace
**Creating input files:**
.. toctree::
:maxdepth: 1
cmfd
element
filter
geometry
material
mesh
nuclide
opencg_compatible
plots
settings
surface
tallies
trigger
universe
**Running OpenMC:**
.. toctree::
:maxdepth: 1
executor
**Post-processing:**
.. toctree::
:maxdepth: 1
particle_restart
statepoint
summary
tallies
**Example Jupyter Notebooks:**
.. toctree::
:maxdepth: 1
examples/pandas-dataframes
examples/tally-arithmetic
.. _Jupyter: https://jupyter.org/
.. _NumPy: http://www.numpy.org/
.. _Codecademy: https://www.codecademy.com/tracks/python
.. _Scipy lectures: https://scipy-lectures.github.io/

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@ -0,0 +1,8 @@
.. _pythonapi_material:
=========
Materials
=========
.. automodule:: openmc.material
:members:

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@ -0,0 +1,8 @@
.. _pythonapi_mesh:
====
Mesh
====
.. automodule:: openmc.mesh
:members:

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@ -0,0 +1,8 @@
.. _pythonapi_nuclide:
=======
Nuclide
=======
.. automodule:: openmc.nuclide
:members:

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@ -0,0 +1,8 @@
.. _pythonapi_opencg_compatible:
====================
OpenCG Compatibility
====================
.. automodule:: openmc.opencg_compatible
:members:

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@ -0,0 +1,8 @@
.. _pythonapi_particle_restart:
================
Particle Restart
================
.. automodule:: openmc.particle_restart
:members:

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@ -0,0 +1,8 @@
.. _pythonapi_plots:
=====
Plots
=====
.. automodule:: openmc.plots
:members:

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@ -0,0 +1,8 @@
.. _pythonapi_settings:
========
Settings
========
.. automodule:: openmc.settings
:members:

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@ -0,0 +1,8 @@
.. _pythonapi_statepoint:
==========
Statepoint
==========
.. automodule:: openmc.statepoint
:members:

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@ -0,0 +1,8 @@
.. _pythonapi_summary:
=======
Summary
=======
.. automodule:: openmc.summary
:members:

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@ -0,0 +1,8 @@
.. _pythonapi_surface:
=======
Surface
=======
.. automodule:: openmc.surface
:members:

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@ -0,0 +1,8 @@
.. _pythonapi_tallies:
=======
Tallies
=======
.. automodule:: openmc.tallies
:members:

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@ -0,0 +1,8 @@
.. _pythonapi_trigger:
=======
Trigger
=======
.. automodule:: openmc.trigger
:members:

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@ -0,0 +1,8 @@
.. _pythonapi_universe:
========
Universe
========
.. automodule:: openmc.universe
:members:

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@ -22,14 +22,13 @@ package manager`_. Simply enter the following commands into the terminal:
sudo apt-get update
sudo apt-get install openmc
Currently, the binary package does not allow for parallel simulations, HDF5_, or
CMFD acceleration through PETSc_. Users who need such capabilities should build
OpenMC from source as is described in :ref:`usersguide_install`.
Currently, the binary package does not allow for parallel simulations or use of
HDF5_. Users who need such capabilities should build OpenMC from source as is
described in :ref:`usersguide_install`.
.. _Personal Package Archive: https://launchpad.net/~paulromano/+archive/staging
.. _APT package manager: https://help.ubuntu.com/community/AptGet/Howto
.. _HDF5: http://www.hdfgroup.org/HDF5/
.. _PETSc: http://www.mcs.anl.gov/petsc/
-------------------------------------------
Installing from Source on Linux or Mac OS X
@ -41,19 +40,21 @@ the following commands in a terminal:
.. code-block:: sh
git clone git://github.com/mit-crpg/openmc.git
cd openmc/src
git clone https://github.com/mit-crpg/openmc.git
cd openmc
git checkout -b master origin/master
mkdir build && cd build
cmake ..
make
sudo make install
This will build an executable named ``openmc`` and install it (by default in
/usr/local/bin). If you do not have administrator privileges, the last command
can be replaced with a local install, e.g.
/usr/local/bin). If you do not have administrator privileges, the cmake command
should specify an installation directory where you have write access, e.g.
.. code-block:: sh
make install -e prefix=$HOME/.local
cmake -DCMAKE_INSTALL_PREFIX=$HOME/.local ..
.. _GitHub: https://github.com/mit-crpg/openmc
.. _git: http://git-scm.com

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.. _releasenotes:
==============================
Release Notes for OpenMC 0.7.0
==============================
-------------------
System Requirements
-------------------
There are no special requirements for running the OpenMC code. As of this
release, OpenMC has been tested on a variety of Linux distributions, Mac OS X,
and Microsoft Windows 7. Memory requirements will vary depending on the size of
the problem at hand (mostly on the number of nuclides in the problem).
------------
New Features
------------
- Complete Python API
- Python 3 compatability for all scripts
- All scripts consistently named openmc-* and installed together
- New 'distribcell' tally filter for repeated cells
- Ability to specify outer lattice universe
- XML input validation utility (openmc-validate-xml)
- Support for hexagonal lattices
- Material union energy grid method
- Tally triggers
- Remove dependence on PETSc
- Significant OpenMP performance improvements
- Support for Fortran 2008 MPI interface
- Use of Travis CI for continuous integration
- Simplifications and improvements to test suite
---------
Bug Fixes
---------
- b5f712_: Fix bug in spherical harmonics tallies
- e6675b_: Ensure all constants are double precision
- 04e2c1_: Fix potential bug in sample_nuclide routine
- 6121d9_: Fix bugs related to particle track files
- 2f0e89_: Fixes for nuclide specification in tallies
.. _b5f712: https://github.com/mit-crpg/openmc/commit/b5f712
.. _e6675b: https://github.com/mit-crpg/openmc/commit/e6675b
.. _04e2c1: https://github.com/mit-crpg/openmc/commit/04e2c1
.. _6121d9: https://github.com/mit-crpg/openmc/commit/6121d9
.. _2f0e89: https://github.com/mit-crpg/openmc/commit/2f0e89
------------
Contributors
------------
This release contains new contributions from the following people:
- `Will Boyd <wbinventor@gmail.com>`_
- `Matt Ellis <mellis13@mit.edu>`_
- `Sterling Harper <sterlingmharper@mit.edu>`_
- `Bryan Herman <bherman@mit.edu>`_
- `Nicholas Horelik <nicholas.horelik@gmail.com>`_
- `Colin Josey <cjosey@mit.edu>`_
- `William Lyu <PaleNeutron@users.noreply.github.com>`_
- `Adam Nelson <nelsonag@umich.edu>`_
- `Paul Romano <paul.k.romano@gmail.com>`_
- `Anthony Scopatz <scopatz@gmail.com>`_
- `Jon Walsh <walshjon@mit.edu>`_

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@ -1,25 +0,0 @@
.. _releasenotes:
=============
Release Notes
=============
The release notes for OpenMC give a list of system requirements, new features,
bugs fixed, and known issues for each successive release.
.. toctree::
:maxdepth: 1
notes_0.6.2
notes_0.6.1
notes_0.6.0
notes_0.5.4
notes_0.5.3
notes_0.5.2
notes_0.5.1
notes_0.5.0
notes_0.4.4
notes_0.4.3
notes_0.4.2
notes_0.4.1
notes_0.4.0

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@ -1,36 +0,0 @@
.. _notes_0.4.0:
==============================
Release Notes for OpenMC 0.4.0
==============================
-------------------
System Requirements
-------------------
There are no special requirements for running the OpenMC code. As of this
release, OpenMC has been tested on a variety of Linux distributions as well as
Mac OS X. However, it has not been tested yet on any releases of Microsoft
Windows. Memory requirements will vary depending on the size of the problem at
hand (mostly on the number of nuclides in the problem).
------------
New Features
------------
- The probability table method for treatment of energy self-shielding in the
unresolved resonance range has been implemented and is now turned on by
default.
- Calculation of Shannon entropy for assessing convergence of the fission source
distribution.
- Ability to compile with the PGI Fortran compiler.
- Ability to run on IBM BlueGene/P machines.
- Completely rewrote how nested universes are handled. Geometry is now much more
robust.
---------
Bug Fixes
---------
- Many geometry errors have been fixed. The Monte Carlo performance benchmark
can now be successfully run in OpenMC.

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@ -1,55 +0,0 @@
.. _notes_0.4.1:
==============================
Release Notes for OpenMC 0.4.1
==============================
-------------------
System Requirements
-------------------
There are no special requirements for running the OpenMC code. As of this
release, OpenMC has been tested on a variety of Linux distributions as well as
Mac OS X. However, it has not been tested yet on any releases of Microsoft
Windows. Memory requirements will vary depending on the size of the problem at
hand (mostly on the number of nuclides in the problem).
------------
New Features
------------
- A batching method has been implemented so that statistics can be calculated
based on multiple generations instead of a single generation. This can help to
overcome problems with underpredicted variance in problems where there is
correlation between successive fission source iterations.
- Users now have the option to select a non-unionized energy grid for problems
with many nuclides where the use of a unionized grid is not feasible.
- Improved plotting capability (Nick Horelik). The plotting input is now in
``plots.xml`` instead of ``plot.xml``.
- Added multiple estimators for k-effective and added a global tally for
leakage.
- Moved cross section-related output into cross_sections.out.
- Improved timing capabilities.
- Can now use more than 2**31 - 1 particles per generation.
- Improved fission bank synchronization method. This also necessitated changing
the source bank to be of type Bank rather than of type Particle.
- Added HDF5 output (not complete yet).
- Major changes to tally implementation.
---------
Bug Fixes
---------
- `b206a8`_: Fixed subtle error in the sampling of energy distributions.
- `800742`_: Fixed error in sampling of angle and rotating angles.
- `a07c08`_: Fixed bug in linear-linear interpolation during sampling energy.
- `a75283`_: Fixed energy and energyout tally filters to support many bins.
- `95cfac`_: Fixed error in cell neighbor searches.
- `83a803`_: Fixed bug related to probability tables.
.. _b206a8: https://github.com/mit-crpg/openmc/commit/b206a8
.. _800742: https://github.com/mit-crpg/openmc/commit/800742
.. _a07c08: https://github.com/mit-crpg/openmc/commit/a07c08
.. _a75283: https://github.com/mit-crpg/openmc/commit/a75283
.. _95cfac: https://github.com/mit-crpg/openmc/commit/95cfac
.. _83a803: https://github.com/mit-crpg/openmc/commit/83a803

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@ -1,56 +0,0 @@
.. _notes_0.4.2:
==============================
Release Notes for OpenMC 0.4.2
==============================
-------------------
System Requirements
-------------------
There are no special requirements for running the OpenMC code. As of this
release, OpenMC has been tested on a variety of Linux distributions, Mac OS X,
and Microsoft Windows 7. Memory requirements will vary depending on the size of
the problem at hand (mostly on the number of nuclides in the problem).
------------
New Features
------------
- Ability to specify void materials.
- Option to not reduce tallies across processors at end of each batch.
- Uniform fission site method for reducing variance on local tallies.
- Reading/writing binary source files.
- Added more messages for <trace> or high verbosity.
- Estimator for diffusion coefficient.
- Ability to specify 'point' source type.
- Ability to change random number seed.
- Users can now specify units='sum' on a <density> tag. This tells the code that
the total material density is the sum of the atom fractions listed for each
nuclide on the material.
---------
Bug Fixes
---------
- a27f8f_: Fixed runtime error bug when using Intel compiler with DEBUG on.
- afe121_: Fixed minor bug in fission bank algorithms.
- e0968e_: Force re-evaluation of cross-sections when each particle is born.
- 298db8_: Fixed bug in surface currents when using energy-in filter.
- 2f3bbe_: Fixed subtle bug in S(a,b) cross section calculation.
- 671f30_: Fixed surface currents on mesh not encompassing geometry.
- b2c40e_: Fixed bug in incoming energy filter for track-length tallies.
- 5524fd_: Mesh filter now works with track-length tallies.
- d050c7_: Added Bessel's correction to make estimate of variance unbiased.
- 2a5b9c_: Fixed regression in plotting.
.. _a27f8f: https://github.com/mit-crpg/openmc/commit/a27f8f
.. _afe121: https://github.com/mit-crpg/openmc/commit/afe121
.. _e0968e: https://github.com/mit-crpg/openmc/commit/e0968e
.. _298db8: https://github.com/mit-crpg/openmc/commit/298db8
.. _2f3bbe: https://github.com/mit-crpg/openmc/commit/2f3bbe
.. _671f30: https://github.com/mit-crpg/openmc/commit/671f30
.. _b2c40e: https://github.com/mit-crpg/openmc/commit/b2c40e
.. _5524fd: https://github.com/mit-crpg/openmc/commit/5524fd
.. _d050c7: https://github.com/mit-crpg/openmc/commit/d050c7
.. _2a5b9c: https://github.com/mit-crpg/openmc/commit/2a5b9c

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@ -1,53 +0,0 @@
.. _notes_0.4.3:
==============================
Release Notes for OpenMC 0.4.3
==============================
-------------------
System Requirements
-------------------
There are no special requirements for running the OpenMC code. As of this
release, OpenMC has been tested on a variety of Linux distributions, Mac OS X,
and Microsoft Windows 7. Memory requirements will vary depending on the size of
the problem at hand (mostly on the number of nuclides in the problem).
------------
New Features
------------
- Option to report confidence intervals for tally results.
- Rotation and translation for filled cells.
- Ability to explicitly specify <estimator> for tallies.
- Ability to store state points and use them to restart runs.
- Fixed source calculations (no subcritical multiplication however).
- Expanded options for external source distribution.
- Ability to tally reaction rates for individual nuclides within a material.
- Reduced memory usage by removing redundant storage or some cross-sections.
- 3bd35b_: Log-log interpolation for URR probability tables.
- Support to specify labels on tallies (nelsonag_).
---------
Bug Fixes
---------
- 33f29a_: Handle negative values in probability table.
- 1c472d_: Fixed survival biasing with probability tables.
- 3c6e80_: Fixed writing tallies with no filters.
- 460ef1_: Invalid results for duplicate tallies.
- 0069d5_: Fixed bug with 0 inactive batches.
- 7af2cf_: Fixed bug in score_analog_tallies.
- 85a60e_: Pick closest angular distribution for law 61.
- 3212f5_: Fixed issue with blank line at beginning of XML files.
.. _nelsonag: https://github.com/nelsonag
.. _33f29a: https://github.com/mit-crpg/openmc/commit/33f29a
.. _1c472d: https://github.com/mit-crpg/openmc/commit/1c472d
.. _3c6e80: https://github.com/mit-crpg/openmc/commit/3c6e80
.. _3bd35b: https://github.com/mit-crpg/openmc/commit/3bd35b
.. _0069d5: https://github.com/mit-crpg/openmc/commit/0069d5
.. _7af2cf: https://github.com/mit-crpg/openmc/commit/7af2cf
.. _460ef1: https://github.com/mit-crpg/openmc/commit/460ef1
.. _85a60e: https://github.com/mit-crpg/openmc/commit/85a60e
.. _3212f5: https://github.com/mit-crpg/openmc/commit/3212f5

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@ -1,45 +0,0 @@
.. _notes_0.4.4:
==============================
Release Notes for OpenMC 0.4.4
==============================
-------------------
System Requirements
-------------------
There are no special requirements for running the OpenMC code. As of this
release, OpenMC has been tested on a variety of Linux distributions, Mac OS X,
and Microsoft Windows 7. Memory requirements will vary depending on the size of
the problem at hand (mostly on the number of nuclides in the problem).
------------
New Features
------------
- Ability to write state points when using <no_reduce>.
- Real-time XML validation in GNU Emacs with RELAX NG schemata.
- Writing state points every n batches with <state_point interval="..." />
- Suppress creation of summary.out and cross_sections.out by default with option
to turn them on with <output> tag in settings.xml file.
- Ability to create HDF5 state points.
- Binary source file is now part of state point file by default.
- Enhanced state point usage and added state point Python scripts.
- Turning confidence intervals on affects k-effective.
- Option to specify <upper_right> for tally meshes.
---------
Bug Fixes
---------
- 4654ee_: Fixed plotting with void cells.
- 7ee461_: Fixed bug with multi-line input using type='word'.
- 792eb3_: Fixed degrees of freedom for confidence intervals.
- 7fd617_: Fixed bug with restart runs in parallel.
- dc4a8f_: Fixed bug with fixed source restart runs.
.. _4654ee: https://github.com/mit-crpg/openmc/commit/4654ee
.. _7ee461: https://github.com/mit-crpg/openmc/commit/7ee461
.. _792eb3: https://github.com/mit-crpg/openmc/commit/792eb3
.. _7fd617: https://github.com/mit-crpg/openmc/commit/7fd617
.. _dc4a8f: https://github.com/mit-crpg/openmc/commit/dc4a8f

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@ -1,52 +0,0 @@
.. _notes_0.5.0:
==============================
Release Notes for OpenMC 0.5.0
==============================
-------------------
System Requirements
-------------------
There are no special requirements for running the OpenMC code. As of this
release, OpenMC has been tested on a variety of Linux distributions, Mac OS X,
and Microsoft Windows 7. Memory requirements will vary depending on the size of
the problem at hand (mostly on the number of nuclides in the problem).
------------
New Features
------------
- All user input options that formerly accepted "off" or "on" should now be
"false" or "true" (the proper XML schema datatype).
- The <criticality> element is deprecated and was replaced with <eigenvalue>.
- Added 'events' score that returns number of events that scored to a tally.
- Restructured tally filter implementation and user input.
- Source convergence acceleration via CMFD (implemented with PETSc).
- Ability to read source files in parallel when number of particles is greater
than that number of source sites.
- Cone surface types.
---------
Bug Fixes
---------
- 737b90_: Coincident surfaces from separate universes / particle traveling
tangent to a surface.
- a819b4_: Output of surface neighbors in summary.out file.
- b11696_: Reading long attribute lists in XML input.
- 2bd46a_: Search for tallying nuclides when no default_xs specified.
- 7a1f08_: Fix word wrapping when writing messages.
- c0e3ec_: Prevent underflow when compiling with MPI=yes and DEBUG=yes.
- 6f8d9d_: Set default tally labels.
- 6a3a5e_: Fix problem with corner-crossing in lattices.
.. _737b90: https://github.com/mit-crpg/openmc/commit/737b90
.. _a819b4: https://github.com/mit-crpg/openmc/commit/a819b4
.. _b11696: https://github.com/mit-crpg/openmc/commit/b11696
.. _2bd46a: https://github.com/mit-crpg/openmc/commit/2bd46a
.. _7a1f08: https://github.com/mit-crpg/openmc/commit/7a1f08
.. _c0e3ec: https://github.com/mit-crpg/openmc/commit/c0e3ec
.. _6f8d9d: https://github.com/mit-crpg/openmc/commit/6f8d9d
.. _6a3a5e: https://github.com/mit-crpg/openmc/commit/6a3a5e

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@ -1,45 +0,0 @@
.. _notes_0.5.1:
==============================
Release Notes for OpenMC 0.5.1
==============================
-------------------
System Requirements
-------------------
There are no special requirements for running the OpenMC code. As of this
release, OpenMC has been tested on a variety of Linux distributions, Mac OS X,
and Microsoft Windows 7. Memory requirements will vary depending on the size of
the problem at hand (mostly on the number of nuclides in the problem).
------------
New Features
------------
- Absorption and combined estimators for k-effective.
- Natural elements can now be specified in materials using <element> rather than
<nuclide>.
- Support for multiple S(a,b) tables in a single material (e.g. BeO).
- Test suite using Python nosetests.
- Proper install capability with 'make install'.
- Lattices can now be 2 or 3 dimensions.
- New scatter-PN score type.
- New kappa-fission score type.
- Ability to tally any reaction by specifying MT.
---------
Bug Fixes
---------
- 94103e_: Two checks for outgoing energy filters.
- e77059_: Fix reaction name for MT=849.
- b0fe88_: Fix distance to surface for cones.
- 63bfd2_: Fix tracklength tallies with cell filter and universes.
- 88daf7_: Fix analog tallies with survival biasing.
.. _94103e: https://github.com/mit-crpg/openmc/commit/94103e
.. _e77059: https://github.com/mit-crpg/openmc/commit/e77059
.. _b0fe88: https://github.com/mit-crpg/openmc/commit/b0fe88
.. _63bfd2: https://github.com/mit-crpg/openmc/commit/63bfd2
.. _88daf7: https://github.com/mit-crpg/openmc/commit/88daf7

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@ -1,57 +0,0 @@
.. _notes_0.5.2:
==============================
Release Notes for OpenMC 0.5.2
==============================
-------------------
System Requirements
-------------------
There are no special requirements for running the OpenMC code. As of this
release, OpenMC has been tested on a variety of Linux distributions, Mac OS X,
and Microsoft Windows 7. Memory requirements will vary depending on the size of
the problem at hand (mostly on the number of nuclides in the problem).
------------
New Features
------------
- Python script for mesh tally plotting
- Isotopic abundances based on IUPAC 2009 when using <element>
- Particle restart files for debugging
- Code will abort after certain number of lost particles (defaults to 10)
- Region outside lattice can be filled with material (void by default)
- 3D voxel plots
- Full HDF5/PHDF5 support (including support in statepoint.py)
- Cell overlap checking with -g command line flag (or when plotting)
---------
Bug Fixes
---------
- 7632f3_: Fixed bug in statepoint.py for multiple generations per batch.
- f85ac4_: Fix infinite loop bug in error module.
- 49c36b_: Don't convert surface ids if surface filter is for current tallies.
- 5ccc78_: Fix bug in reassignment of bins for mesh filter.
- b1f52f_: Fixed bug in plot color specification.
- eae7e5_: Fixed many memory leaks.
- 10c1cc_: Minor CMFD fixes.
- afdb50_: Add compatibility for XML comments without whitespace.
- a3c593_: Fixed bug in use of free gas scattering for H-1.
- 3a66e3_: Fixed bug in 2D mesh tally implementation.
- ab0793_: Corrected PETSC_NULL references to their correct types.
- 182ebd_: Use analog estimator with energyout filter.
.. _7632f3: https://github.com/mit-crpg/openmc/commit/7632f3
.. _f85ac4: https://github.com/mit-crpg/openmc/commit/f85ac4
.. _49c36b: https://github.com/mit-crpg/openmc/commit/49c36b
.. _5ccc78: https://github.com/mit-crpg/openmc/commit/5ccc78
.. _b1f52f: https://github.com/mit-crpg/openmc/commit/b1f52f
.. _eae7e5: https://github.com/mit-crpg/openmc/commit/eae7e5
.. _10c1cc: https://github.com/mit-crpg/openmc/commit/10c1cc
.. _afdb50: https://github.com/mit-crpg/openmc/commit/afdb50
.. _a3c593: https://github.com/mit-crpg/openmc/commit/a3c593
.. _3a66e3: https://github.com/mit-crpg/openmc/commit/3a66e3
.. _ab0793: https://github.com/mit-crpg/openmc/commit/ab0793
.. _182ebd: https://github.com/mit-crpg/openmc/commit/182ebd

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@ -1,49 +0,0 @@
.. _notes_0.5.3:
==============================
Release Notes for OpenMC 0.5.3
==============================
-------------------
System Requirements
-------------------
There are no special requirements for running the OpenMC code. As of this
release, OpenMC has been tested on a variety of Linux distributions, Mac OS X,
and Microsoft Windows 7. Memory requirements will vary depending on the size of
the problem at hand (mostly on the number of nuclides in the problem).
------------
New Features
------------
- Output interface enhanced to allow multiple files handles to be opened
- Particle restart file linked to output interface
- Particle restarts and state point restarts are both identified with the -r
command line flag.
- Particle instance no longer global, passed to all physics routines
- Physics routines refactored to rely less on global memory, more arguments
passed in
- CMFD routines refactored and now can compute dominance ratio on the fly
- PETSc 3.4.2 or higher must be used and compiled with fortran datatype support
- Memory leaks fixed except for ones from xml-fortran package
- Test suite enhanced to test output with different compiler options
- Description of OpenMC development workflow added
- OpenMP shared-memory parallelism added
- Special run mode --tallies removed.
---------
Bug Fixes
---------
- 2b1e8a_: Normalize direction vector after reflecting particle.
- 5853d2_: Set blank default for cross section listing alias.
- e178c7_: Fix infinite loop with words greater than 80 characters in write_message.
- c18a6e_: Check for valid secondary mode on S(a,b) tables.
- 82c456_: Fix bug where last process could have zero particles.
.. _2b1e8a: https://github.com/mit-crpg/openmc/commit/2b1e8a
.. _5853d2: https://github.com/mit-crpg/openmc/commit/5853d2
.. _e178c7: https://github.com/mit-crpg/openmc/commit/e178c7
.. _c18a6e: https://github.com/mit-crpg/openmc/commit/c18a6e
.. _82c456: https://github.com/mit-crpg/openmc/commit/82c456

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@ -1,64 +0,0 @@
.. _notes_0.5.4:
==============================
Release Notes for OpenMC 0.5.4
==============================
-------------------
System Requirements
-------------------
There are no special requirements for running the OpenMC code. As of this
release, OpenMC has been tested on a variety of Linux distributions, Mac OS X,
and Microsoft Windows 7. Memory requirements will vary depending on the size of
the problem at hand (mostly on the number of nuclides in the problem).
------------
New Features
------------
- Source sites outside geometry are resampled
- XML-Fortran backend replaced by FoX XML
- Ability to write particle track files
- Handle lost particles more gracefully (via particle track files)
- Multiple random number generator streams
- Mesh tally plotting utility converted to use Tkinter rather than PyQt
- Script added to download ACE data from NNDC
- Mixed ASCII/binary cross_sections.xml now allowed
- Expanded options for writing source bank
- Re-enabled ability to use source file as starting source
- S(a,b) recalculation avoided when same nuclide and S(a,b) table are accessed
---------
Bug Fixes
---------
- 32c03c_: Check for valid data in cross_sections.xml
- c71ef5_: Fix bug in statepoint.py
- 8884fb_: Check for all ZAIDs for S(a,b) tables
- b38af0_: Fix XML reading on multiple levels of input
- d28750_: Fix bug in convert_xsdir.py
- cf567c_: ENDF/B-VI data checked for compatibility
- 6b9461_: Fix p_valid sampling inside of sample_energy
.. _32c03c: https://github.com/mit-crpg/openmc/commit/32c03c
.. _c71ef5: https://github.com/mit-crpg/openmc/commit/c71ef5
.. _8884fb: https://github.com/mit-crpg/openmc/commit/8884fb
.. _b38af0: https://github.com/mit-crpg/openmc/commit/b38af0
.. _d28750: https://github.com/mit-crpg/openmc/commit/d28750
.. _cf567c: https://github.com/mit-crpg/openmc/commit/cf567c
.. _6b9461: https://github.com/mit-crpg/openmc/commit/6b9461
------------
Contributors
------------
This release contains new contributions from the following people:
- `Sterling Harper <smharper@mit.edu>`_
- `Bryan Herman <bherman@mit.edu>`_
- `Nick Horelik <nhorelik@mit.edu>`_
- `Adam Nelson <nelsonag@umich.edu>`_
- `Paul Romano <paul.k.romano@gmail.com>`_
- `Tuomas Viitanen <tuomas.viitanen@vtt.fi>`_
- `Jon Walsh <walshjon@mit.edu>`_

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@ -1,59 +0,0 @@
.. _notes_0.6.0:
==============================
Release Notes for OpenMC 0.6.0
==============================
-------------------
System Requirements
-------------------
There are no special requirements for running the OpenMC code. As of this
release, OpenMC has been tested on a variety of Linux distributions, Mac OS X,
and Microsoft Windows 7. Memory requirements will vary depending on the size of
the problem at hand (mostly on the number of nuclides in the problem).
------------
New Features
------------
- Legendre and spherical harmonic expansion tally scores
- CMake is now default build system
- Regression test suite based on CTests and NNDC cross sections
- FoX is now a git submodule
- Support for older cross sections (e.g. MCNP 66c)
- Progress bar for plots
- Expanded support for natural elements via <natural_elements> in settings.xml
---------
Bug Fixes
---------
- 41f7ca_: Fixed erroneous results from survival biasing
- 038736_: Fix tallies over void materials
- 46f9e8_: Check for negative values in probability tables
- d1ca35_: Fixed sampling of angular distribution
- 0291c0_: Fixed indexing error in plotting
- d7a7d0_: Fix bug with <element> specifying xs attribute
- 85b3cb_: Fix out-of-bounds error with OpenMP threading
.. _41f7ca: https://github.com/mit-crpg/openmc/commit/41f7ca
.. _038736: https://github.com/mit-crpg/openmc/commit/038736
.. _46f9e8: https://github.com/mit-crpg/openmc/commit/46f9e8
.. _d1ca35: https://github.com/mit-crpg/openmc/commit/d1ca35
.. _0291c0: https://github.com/mit-crpg/openmc/commit/0291c0
.. _d7a7d0: https://github.com/mit-crpg/openmc/commit/d7a7d0
.. _85b3cb: https://github.com/mit-crpg/openmc/commit/85b3cb
------------
Contributors
------------
This release contains new contributions from the following people:
- `Sterling Harper <smharper@mit.edu>`_
- `Bryan Herman <bherman@mit.edu>`_
- `Nick Horelik <nhorelik@mit.edu>`_
- `Adam Nelson <nelsonag@umich.edu>`_
- `Paul Romano <paul.k.romano@gmail.com>`_
- `Jon Walsh <walshjon@mit.edu>`_

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@ -1,65 +0,0 @@
.. _notes_0.6.1:
==============================
Release Notes for OpenMC 0.6.1
==============================
-------------------
System Requirements
-------------------
There are no special requirements for running the OpenMC code. As of this
release, OpenMC has been tested on a variety of Linux distributions, Mac OS X,
and Microsoft Windows 7. Memory requirements will vary depending on the size of
the problem at hand (mostly on the number of nuclides in the problem).
------------
New Features
------------
- Coarse mesh finite difference (CMFD) acceleration no longer requires PETSc
- Statepoint file numbering is now zero-padded
- Python scripts now compatible with Python 2 or 3
- Ability to run particle restarts in fixed source calculations
- Capability to filter box source by fissionable materials
- Nuclide/element names are now case insensitive in input files
- Improved treatment of resonance scattering for heavy nuclides
---------
Bug Fixes
---------
- 03e890_: Check for energy-dependent multiplicities in ACE files
- 4439de_: Fix distance-to-surface calculation for general plane surface
- 5808ed_: Account for differences in URR band probabilities at different energies
- 2e60c0_: Allow zero atom/weight percents in materials
- 3e0870_: Don't use PWD environment variable when setting path to input files
- dc4776_: Handle probability table resampling correctly
- 01178b_: Fix metastables nuclides in NNDC cross_sections.xml file
- 62ec43_: Don't read tallies.xml when OpenMC is run in plotting mode
- 2a95ef_: Prevent segmentation fault on "current" score without mesh filter
- 93e482_: Check for negative values in probability tables
.. _03e890: https://github.com/mit-crpg/openmc/commit/03e890
.. _4439de: https://github.com/mit-crpg/openmc/commit/4439de
.. _5808ed: https://github.com/mit-crpg/openmc/commit/5808ed
.. _2e60c0: https://github.com/mit-crpg/openmc/commit/2e60c0
.. _3e0870: https://github.com/mit-crpg/openmc/commit/3e0870
.. _dc4776: https://github.com/mit-crpg/openmc/commit/dc4776
.. _01178b: https://github.com/mit-crpg/openmc/commit/01178b
.. _62ec43: https://github.com/mit-crpg/openmc/commit/62ec43
.. _2a95ef: https://github.com/mit-crpg/openmc/commit/2a95ef
.. _93e482: https://github.com/mit-crpg/openmc/commit/93e482
------------
Contributors
------------
This release contains new contributions from the following people:
- `Sterling Harper <smharper@mit.edu>`_
- `Bryan Herman <bherman@mit.edu>`_
- `Adam Nelson <nelsonag@umich.edu>`_
- `Paul Romano <paul.k.romano@gmail.com>`_
- `Jon Walsh <walshjon@mit.edu>`_
- `Will Boyd <wbinventor@gmail.com>`_

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@ -1,58 +0,0 @@
.. _notes_0.6.2:
==============================
Release Notes for OpenMC 0.6.2
==============================
-------------------
System Requirements
-------------------
There are no special requirements for running the OpenMC code. As of this
release, OpenMC has been tested on a variety of Linux distributions, Mac OS X,
and Microsoft Windows 7. Memory requirements will vary depending on the size of
the problem at hand (mostly on the number of nuclides in the problem).
------------
New Features
------------
- Meshline plotting capability
- Support for plotting cells/materials on middle universe levels
- Ability to model cells with no surfaces
- Compatibility with PETSc 3.5
- Compatability with OpenMPI 1.7/1.8
- Improved overall performance via logarithmic-mapped energy grid search
- Improved multi-threaded performance with atomic operations
- Support for fixed source problems with fissionable materials
---------
Bug Fixes
---------
- 26fb93_: Fix problem with -t, --track command-line flag
- 2f07c0_: Improved evaporation spectrum algorithm
- e6abb9_: Fix segfault when tallying in a void material
- 291b45_: Handle metastable nuclides in NNDC data and multiplicities in MT=5 data
.. _26fb93: https://github.com/mit-crpg/openmc/commit/26fb93
.. _2f07c0: https://github.com/mit-crpg/openmc/commit/2f07c0
.. _e6abb9: https://github.com/mit-crpg/openmc/commit/e6abb9
.. _291b45: https://github.com/mit-crpg/openmc/commit/291b45
------------
Contributors
------------
This release contains new contributions from the following people:
- `Will Boyd <wbinventor@gmail.com>`_
- `Matt Ellis <mellis13@mit.edu>`_
- `Sterling Harper <smharper@mit.edu>`_
- `Bryan Herman <bherman@mit.edu>`_
- `Nicholas Horelik <nicholas.horelik@gmail.com>`_
- `Anton Leontiev <bunder@t-25.ru>`_
- `Adam Nelson <nelsonag@umich.edu>`_
- `Paul Romano <paul.k.romano@gmail.com>`_
- `Jon Walsh <walshjon@mit.edu>`_
- `John Xia <john.danger.xia@gmail.com>`_

7
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@ -146,13 +146,12 @@ and `Volume II`_. You may also find it helpful to review the following terms:
.. _constructive solid geometry: http://en.wikipedia.org/wiki/Constructive_solid_geometry
.. _git: http://git-scm.com/
.. _git tutorials: http://git-scm.com/documentation
.. _Reactor Concepts Manual: http://web.mit.edu/romano7/www/reactor_concepts.pdf
.. _Volume I: http://www.hss.doe.gov/nuclearsafety/techstds/docs/handbook/h1019v1.pdf
.. _Volume II: http://www.hss.doe.gov/nuclearsafety/techstds/docs/handbook/h1019v2.pdf
.. _Reactor Concepts Manual: http://www.tayloredge.com/periodic/trivia/ReactorConcepts.pdf
.. _Volume I: http://energy.gov/sites/prod/files/2013/06/f2/h1019v1.pdf
.. _Volume II: http://energy.gov/sites/prod/files/2013/06/f2/h1019v2.pdf
.. _OpenMC source code: https://github.com/mit-crpg/openmc
.. _GitHub: https://github.com/
.. _bug reports: https://github.com/mit-crpg/openmc/issues
.. _Neutron cross section: http://en.wikipedia.org/wiki/Neutron_cross_section
.. _Effective multiplication factor: http://en.wikipedia.org/wiki/Effective_multiplication_factor
.. _Flux: http://en.wikipedia.org/wiki/Neutron_flux

546
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@ -51,6 +51,43 @@ files are called:
* ``plots.xml``
* ``cmfd.xml``
--------------------
Validating XML Files
--------------------
Input files can be checked before executing OpenMC using the
``openmc-validate-xml`` script which is installed alongside the Python API. Two
command line arguments can be set when running ``openmc-validate-xml``:
* ``-i``, ``--input-path`` - Location of OpenMC input files.
*Default*: current working directory
* ``-r``, ``--relaxng-path`` - Location of OpenMC RelaxNG files.
*Default*: None
If the RelaxNG path is not set, the script will search for these files because
it expects that the user is either running the script located in the install
directory ``bin`` folder or in ``src/utils``. Once executed, it will match
OpenMC XML files with their RelaxNG schema and check if they are valid. Below
is a table of the messages that will be printed after each file is checked.
======================== ===================================
Message Description
======================== ===================================
[XML ERROR] Cannot parse XML file.
[NO RELAXNG FOUND] No RelaxNG file found for XML file.
[NOT VALID] XML file does not match RelaxNG.
[VALID] XML file matches RelaxNG.
======================== ===================================
As an example, if OpenMC is installed in the directory
``/opt/openmc/0.6.2`` and the current working directory is where
OpenMC XML input files are located, they can be validated using
the following command:
.. code-block:: bash
/opt/openmc/0.6.2/bin/xml_validate
--------------------------------------
Settings Specification -- settings.xml
--------------------------------------
@ -99,6 +136,8 @@ default. This element has the following attributes/sub-elements:
*Default*: 1.0
.. _eigenvalue:
``<eigenvalue>`` Element
------------------------
@ -130,15 +169,47 @@ should be performed. It has the following attributes/sub-elements:
*Default*: None
:keff_trigger:
This tag specifies a precision trigger on the combined :math:`k_{eff}`. The
trigger is a convergence criterion on the uncertainty of the estimated
eigenvalue. It has the following attributes/sub-elements:
:type:
The type of precision trigger. Accepted options are "variance", "std_dev",
and "rel_err".
:variance:
Variance of the batch mean :math:`\sigma^2`
:std_dev:
Standard deviation of the batch mean :math:`\sigma`
:rel_err:
Relative error of the batch mean :math:`\frac{\sigma}{\mu}`
*Default*: None
:threshold:
The precision trigger's convergence criterion for the
combined :math:`k_{eff}`.
*Default*: None
.. note:: See section on the :ref:`trigger` for more information.
``<energy_grid>`` Element
-------------------------
The ``<energy_grid>`` element determines the treatment of the energy grid during
a simulation. The valid options are "nuclide" and "logarithm". Setting this
element to "nuclide" will cause OpenMC to use a nuclide's energy grid when
determining what points to interpolate between for determining cross sections
(i.e. non-unionized energy grid). Setting this element to "logarithm" causes
OpenMC to use a logarithmic mapping technique described in LA-UR-14-24530_.
a simulation. The valid options are "nuclide", "logarithm", and
"material-union". Setting this element to "nuclide" will cause OpenMC to use a
nuclide's energy grid when determining what points to interpolate between for
determining cross sections (i.e. non-unionized energy grid). Setting this
element to "logarithm" causes OpenMC to use a logarithmic mapping technique
described in LA-UR-14-24530_. Setting this element to "material-union" will
cause OpenMC to create energy grids that are unionized material-by-material and
use these grids when determining the energy-cross section pairs to interpolate
cross section values between.
*Default*: logarithm
@ -170,8 +241,8 @@ problem. It has the following attributes/sub-elements:
``<fixed_source>`` Element
--------------------------
The ``<fixed_source>`` element indicates that a fixed source calculation should be
performed. It has the following attributes/sub-elements:
The ``<fixed_source>`` element indicates that a fixed source calculation should
be performed. It has the following attributes/sub-elements:
:batches:
The total number of batches. For fixed source calculations, each batch
@ -252,7 +323,8 @@ out the file and "false" will not.
*Default*: true
.. note:: The tally results will always be written to a binary/HDF5 state point file.
.. note:: The tally results will always be written to a binary/HDF5 state
point file.
``<output_path>`` Element
-------------------------
@ -369,24 +441,22 @@ attributes/sub-elements:
has the following attributes:
:type:
The type of spatial distribution. Valid options are "box" and "point". A
"box" spatial distribution has coordinates sampled uniformly in a
parallelepiped. A "point" spatial distribution has coordinates specified
by a triplet.
The type of spatial distribution. Valid options are "box", "fission", and
"point". A "box" spatial distribution has coordinates sampled uniformly in
a parallelepiped. A "fission" spatial distribution samples locations from
a "box" distribution but only locations in fissionable materials are
accepted. A "point" spatial distribution has coordinates specified by a
triplet.
*Default*: None
:parameters:
For a "box" spatial distribution, ``parameters`` should be given as six
real numbers, the first three of which specify the lower-left corner of a
parallelepiped and the last three of which specify the upper-right
corner. Source sites are sampled uniformly through that parallelepiped.
To filter a "box" spatial distribution by fissionable material, specify
"fission" tag instead of "box". The ``parameters`` should be given as six
real numbers, the first three of which specify the lower-left corner of a
parallelepiped and the last three of which specify the upper-right
corner. Source sites are sampled uniformly through that parallelepiped.
For a "box" or "fission" spatial distribution, ``parameters`` should be
given as six real numbers, the first three of which specify the lower-left
corner of a parallelepiped and the last three of which specify the
upper-right corner. Source sites are sampled uniformly through that
parallelepiped.
For a "point" spatial distribution, ``parameters`` should be given as
three real numbers which specify the (x,y,z) location of an isotropic
@ -409,7 +479,7 @@ attributes/sub-elements:
:parameters:
For an "isotropic" angular distribution, ``parameters`` should not be
specified
specified.
For a "monodirectional" angular distribution, ``parameters`` should be
given as three real numbers which specify the angular cosines with respect
@ -427,7 +497,7 @@ attributes/sub-elements:
"watt", and "maxwell". The "monoenergetic" option produces source sites at
a single energy. The "watt" option produces source sites whose energy is
sampled from a Watt fission spectrum. The "maxwell" option produce source
sites whose energy is sampled from a Maxwell fission spectrum
sites whose energy is sampled from a Maxwell fission spectrum.
*Default*: watt
@ -501,8 +571,8 @@ attributes/sub-elements:
*Default*: None
:separate:
If this element is set to "true", a separate binary source point file will be
written. Otherwise, the source sites will be written in the state point
If this element is set to "true", a separate binary source point file will
be written. Otherwise, the source sites will be written in the state point
directly.
*Default*: false
@ -533,8 +603,6 @@ survival biasing, otherwise known as implicit capture or absorption.
*Default*: false
.. _trace:
``<threads>`` Element
---------------------
@ -543,6 +611,8 @@ a simulation. It has no attributes and accepts a positive integer value.
*Default*: None (Determined by environment variable :envvar:`OMP_NUM_THREADS`)
.. _trace:
``<trace>`` Element
-------------------
@ -557,10 +627,56 @@ integers: the batch number, generation number, and particle number.
``<track>`` Element
-------------------
The ``<track>`` element specifies particles for which OpenMC will output binary files describing particle position at every step of its transport. This element should be followed by triplets of integers. Each triplet describes one particle. The integers in each triplet specify the batch number, generation number, and particle number, respectively.
The ``<track>`` element specifies particles for which OpenMC will output binary
files describing particle position at every step of its transport. This element
should be followed by triplets of integers. Each triplet describes one
particle. The integers in each triplet specify the batch number, generation
number, and particle number, respectively.
*Default*: None
.. _trigger:
``<trigger>`` Element
-------------------------
OpenMC includes tally precision triggers which allow the user to define
uncertainty thresholds on :math:`k_{eff}` in the ``<eigenvalue>`` subelement of
``settings.xml``, and/or tallies in ``tallies.xml``. When using triggers,
OpenMC will run until it completes as many batches as defined by ``<batches>``.
At this point, the uncertainties on all tallied values are computed and
compared with their corresponding trigger thresholds. If any triggers have not
been met, OpenMC will continue until either all trigger thresholds have been
satisfied or ``<max_batches>`` has been reached.
The ``<trigger>`` element provides an active "toggle switch" for tally
precision trigger(s), the maximum number of batches and the batch interval. It
has the following attributes/sub-elements:
:active:
This determines whether or not to use trigger(s). Trigger(s) are used when
this tag is set to "true".
:max_batches:
This describes the maximum number of batches allowed when using trigger(s).
.. note:: When max_batches is set, the number of ``batches`` shown in
``<eigenvalue>`` element represents minimum number of batches to
simulate when using the trigger(s).
:batch_interval:
This tag describes the number of batches in between convergence checks.
OpenMC will check if the trigger has been reached at each batch defined
by ``batch_interval`` after the minimum number of batches is reached.
.. note:: If this tag is not present, the ``batch_interval`` is predicted
dynamically by OpenMC for each convergence check. The predictive
model assumes no correlation between fission sources
distributions from batch-to-batch. This assumption is reasonable
for fixed source and small criticality calculations, but is very
optimistic for highly coupled full-core reactor problems.
``<uniform_fs>`` Element
------------------------
@ -664,6 +780,12 @@ Each ``<surface>`` element can have the following attributes or sub-elements:
*Default*: None
:name:
An optional string name to identify the surface in summary output
files. This string is limited to 52 characters for formatting purposes.
*Default*: ""
:type:
The type of the surfaces. This can be "x-plane", "y-plane", "z-plane",
"plane", "x-cylinder", "y-cylinder", "z-cylinder", or "sphere".
@ -740,10 +862,16 @@ The following quadratic surfaces can be modeled:
Each ``<cell>`` element can have the following attributes or sub-elements:
:id:
A unique integer that can be used to identify the surface.
A unique integer that can be used to identify the cell.
*Default*: None
:name:
An optional string name to identify the cell in summary output files.
This string is limmited to 52 characters for formatting purposes.
*Default*: ""
:universe:
The ``id`` of the universe that this cell is contained in.
@ -802,19 +930,18 @@ Each ``<cell>`` element can have the following attributes or sub-elements:
---------------------
The ``<lattice>`` can be used to represent repeating structures (e.g. fuel pins
in an assembly) or other geometry which naturally fits into a two- or
three-dimensional structured mesh. Each cell within the lattice is filled with a
specified universe. A ``<lattice>`` accepts the following attributes or
sub-elements:
in an assembly) or other geometry which fits onto a rectilinear grid. Each cell
within the lattice is filled with a specified universe. A ``<lattice>`` accepts
the following attributes or sub-elements:
:id:
A unique integer that can be used to identify the surface.
A unique integer that can be used to identify the lattice.
:type:
A string indicating the arrangement of lattice cells. Currently, the only
accepted option is "rectangular".
:name:
An optional string name to identify the lattice in summary output
files. This string is limited to 52 characters for formatting purposes.
*Default*: rectangular
*Default*: ""
:dimension:
Two or three integers representing the number of lattice cells in the x- and
@ -828,22 +955,118 @@ sub-elements:
*Default*: None
:width:
The width of the lattice cell in the x- and y- (and z-) directions.
:pitch:
If the lattice is 3D, then three real numbers that express the distance
between the centers of lattice cells in the x-, y-, and z- directions. If
the lattice is 2D, then omit the third value.
*Default*: None
:outside:
The unique integer identifier of a material that is to be used to fill all
space outside of the lattice. This element is optional.
:outer:
The unique integer identifier of a universe that will be used to fill all
space outside of the lattice. The universe will be tiled repeatedly as if
it were placed in a lattice of infinite size. This element is optional.
*Default*: The region outside the defined lattice is treated as void.
*Default*: An error will be raised if a particle leaves a lattice with no
outer universe.
:universes:
A list of the universe numbers that fill each cell of the lattice.
*Default*: None
Here is an example of a properly defined 2d rectangular lattice:
.. code-block:: xml
<lattice id="10" dimension="3 3" outer="1">
<lower_left> -1.5 -1.5 </lower_left>
<pitch> 1.0 1.0 </pitch>
<universes>
2 2 2
2 1 2
2 2 2
</universes>
</lattice>
``<hex_lattice>`` Element
-------------------------
The ``<hex_lattice>`` can be used to represent repeating structures (e.g. fuel
pins in an assembly) or other geometry which naturally fits onto a hexagonal
grid or hexagonal prism grid. Each cell within the lattice is filled with a
specified universe. This lattice uses the "flat-topped hexagon" scheme where two
of the six edges are perpendicular to the y-axis. A ``<hex_lattice>`` accepts
the following attributes or sub-elements:
:id:
A unique integer that can be used to identify the lattice.
:name:
An optional string name to identify the hex_lattice in summary output
files. This string is limited to 52 characters for formatting purposes.
*Default*: ""
:n_rings:
An integer representing the number of radial ring positions in the xy-plane.
Note that this number includes the degenerate center ring which only has one
element.
*Default*: None
:n_axial:
An integer representing the number of positions along the z-axis. This
element is optional.
*Default*: None
:center:
The coordinates of the center of the lattice. If the lattice does not have
axial sections then only the x- and y-coordinates are specified.
*Default*: None
:pitch:
If the lattice is 3D, then two real numbers that express the distance
between the centers of lattice cells in the xy-plane and along the z-axis,
respectively. If the lattice is 2D, then omit the second value.
*Default*: None
:outer:
The unique integer identifier of a universe that will be used to fill all
space outside of the lattice. The universe will be tiled repeatedly as if
it were placed in a lattice of infinite size. This element is optional.
*Default*: An error will be raised if a particle leaves a lattice with no
outer universe.
:universes:
A list of the universe numbers that fill each cell of the lattice.
*Default*: None
Here is an example of a properly defined 2d hexagonal lattice:
.. code-block:: xml
<hex_lattice id="10" n_rings="3" outer="1">
<center> 0.0 0.0 </center>
<pitch> 1.0 </pitch>
<universes>
202
202 202
202 202 202
202 202
202 101 202
202 202
202 202 202
202 202
202
</universes>
</hex_lattice>
.. _constructive solid geometry: http://en.wikipedia.org/wiki/Constructive_solid_geometry
.. _quadratic surfaces: http://en.wikipedia.org/wiki/Quadric
@ -862,6 +1085,12 @@ Each ``material`` element can have the following attributes or sub-elements:
:id:
A unique integer that can be used to identify the material.
:name:
An optional string name to identify the material in summary output
files. This string is limited to 52 characters for formatting purposes.
*Default*: ""
:density:
An element with attributes/sub-elements called ``value`` and ``units``. The
``value`` attribute is the numeric value of the density while the ``units``
@ -959,15 +1188,18 @@ post-collision energy, and an arbitrary structured mesh.
The three valid elements in the tallies.xml file are ``<tally>``, ``<mesh>``,
and ``<assume_separate>``.
.. _tally:
``<tally>`` Element
-------------------
The ``<tally>`` element accepts the following sub-elements:
:label:
This is an optional sub-element specifying the name of this tally to be used
for output purposes. This string is limited to 52 characters for formatting
purposes.
:name:
An optional string name to identify the tally in summary output
files. This string is limited to 52 characters for formatting purposes.
*Default*: ""
:filter:
Specify a filter that restricts contributions to the tally to particles
@ -981,11 +1213,13 @@ The ``<tally>`` element accepts the following sub-elements:
The ``filter`` element has the following attributes/sub-elements:
:type:
The type of the filter. Accepted options are "cell", "cellborn", "material",
"universe", "energy", "energyout", and "mesh".
The type of the filter. Accepted options are "cell", "cellborn",
"material", "universe", "energy", "energyout", "mesh", and
"distribcell".
:bins:
For each filter type, the corresponding ``bins`` entry is given as follows:
For each filter type, the corresponding ``bins`` entry is given as
follows:
:cell:
A list of cells in which the tally should be accumulated.
@ -1020,6 +1254,15 @@ The ``<tally>`` element accepts the following sub-elements:
:mesh:
The ``id`` of a structured mesh to be tallied over.
:distribcell:
The single cell which should be tallied uniquely for all instances.
.. note::
The distribcell filter will take a single cell ID and will tally
each unique occurrence of that cell separately. This filter will
not accept more than one cell ID. It is not recommended to combine
this filter with a cell or mesh filter.
:nuclides:
If specified, the scores listed will be for particular nuclides, not the
summation of reactions from all nuclides. The format for nuclides should be
@ -1053,24 +1296,25 @@ The ``<tally>`` element accepts the following sub-elements:
physical quantities:
:flux:
Total flux
Total flux in particle-cm per source particle.
:total:
Total reaction rate
Total reaction rate in reactions per source particle.
:scatter:
Total scattering rate. Can also be identified with the ``scatter-0``
response type.
response type. Units are reactions per source particle.
:absorption:
Total absorption rate. This accounts for all reactions which do not
produce secondary neutrons.
produce secondary neutrons. Units are reactions per source particle.
:fission:
Total fission rate
Total fission rate in reactions per source particle.
:nu-fission:
Total production of neutrons due to fission
Total production of neutrons due to fission. Units are neutrons produced
per source neutron.
:kappa-fission:
The recoverable energy production rate due to fission. The recoverable
@ -1079,51 +1323,55 @@ The ``<tally>`` element accepts the following sub-elements:
total energies, and the total energy released by the delayed :math:`\beta`
particles. The neutrino energy does not contribute to this response. The
prompt and delayed :math:`\gamma`-rays are assumed to deposit their energy
locally.
locally. Units are MeV per source particle.
:scatter-N:
Tally the N\ :sup:`th` \ scattering moment, where N is the Legendre
expansion order of the change in particle angle :math:`\left(\mu\right)`.
N must be between 0 and 10. As an example, tallying the
2\ :sup:`nd` \ scattering moment would be specified as
``<scores> scatter-2 </scores>``.
N must be between 0 and 10. As an example, tallying the 2\ :sup:`nd` \
scattering moment would be specified as ``<scores> scatter-2
</scores>``. Units are reactions per source particle.
:scatter-PN:
Tally all of the scattering moments from order 0 to N, where N is the
Legendre expansion order of the change in particle angle :math:`\left(\mu\right)`.
That is, ``scatter-P1`` is equivalent to requesting tallies of
``scatter-0`` and ``scatter-1``. Like for ``scatter-N``,
N must be between 0 and 10. As an example, tallying up to the
2\ :sup:`nd` \ scattering moment would be specified as
``<scores> scatter-P2 </scores>``.
Legendre expansion order of the change in particle angle
:math:`\left(\mu\right)`. That is, ``scatter-P1`` is equivalent to
requesting tallies of ``scatter-0`` and ``scatter-1``. Like for
``scatter-N``, N must be between 0 and 10. As an example, tallying up to
the 2\ :sup:`nd` \ scattering moment would be specified as ``<scores>
scatter-P2 </scores>``. Units are reactions per source particle.
:scatter-YN:
``scatter-YN`` is similar to ``scatter-PN`` except an additional
expansion is performed for the incoming particle direction
:math:`\left(\Omega\right)` using the real spherical harmonics. This is useful
for performing angular flux moment weighting of the scattering moments.
Like ``scatter-PN``, ``scatter-YN`` will tally all of the moments from
order 0 to N; N again must be between 0 and 10.
``scatter-YN`` is similar to ``scatter-PN`` except an additional expansion
is performed for the incoming particle direction
:math:`\left(\Omega\right)` using the real spherical harmonics. This is
useful for performing angular flux moment weighting of the scattering
moments. Like ``scatter-PN``, ``scatter-YN`` will tally all of the moments
from order 0 to N; N again must be between 0 and 10. Units are reactions
per source particle.
:nu-scatter, nu-scatter-N, nu-scatter-PN, nu-scatter-YN:
These scores are similar in functionality to their ``scatter*``
equivalents except the total production of neutrons due to
scattering is scored vice simply the scattering rate. This accounts for
multiplicity from (n,2n), (n,3n), and (n,4n) reactions.
equivalents except the total production of neutrons due to scattering is
scored vice simply the scattering rate. This accounts for multiplicity
from (n,2n), (n,3n), and (n,4n) reactions. Units are neutrons produced per
source particle.
:flux-YN:
Spherical harmonic expansion of the direction of motion
:math:`\left(\Omega\right)` of the total flux. This score will tally
all of the harmonic moments of order 0 to N. N must be between 0 and 10.
:math:`\left(\Omega\right)` of the total flux. This score will tally all
of the harmonic moments of order 0 to N. N must be between 0
and 10. Units are particle-cm per source particle.
:total-YN:
The total reaction rate expanded via spherical harmonics about the
direction of motion of the neutron, :math:`\Omega`.
This score will tally all of the harmonic moments of order 0 to N. N must
be between 0 and 10.
be between 0 and 10. Units are reactions per source particle.
:current:
Partial currents on the boundaries of each cell in a mesh.
Partial currents on the boundaries of each cell in a mesh. Units are
particles per source particle.
.. note::
This score can only be used if a mesh filter has been
@ -1131,7 +1379,41 @@ The ``<tally>`` element accepts the following sub-elements:
other score.
:events:
Number of scoring events
Number of scoring events. Units are events per source particle.
:trigger:
Precision trigger applied to all filter bins and nuclides for this tally.
It must specify the trigger's type, threshold and scores to which it will
be applied. It has the following attributes/sub-elements:
:type:
The type of the trigger. Accepted options are "variance", "std_dev",
and "rel_err".
:variance:
Variance of the batch mean :math:`\sigma^2`
:std_dev:
Standard deviation of the batch mean :math:`\sigma`
:rel_err:
Relative error of the batch mean :math:`\frac{\sigma}{\mu}`
*Default*: None
:threshold:
The precision trigger's convergence criterion for tallied values.
*Default*: None
:scores:
The score(s) in this tally to which the trigger should be applied.
.. note:: The ``scores`` in ``trigger`` must have been defined in
``scores`` in ``tally``. An optional "all" may be used to
select all scores in this tally.
*Default*: "all"
``<mesh>`` Element
------------------
@ -1171,8 +1453,8 @@ overhead. The effect of assuming all tallies are spatially separate is that once
one tally is scored to, the same event is assumed not to score to any other
tallies. This element should be followed by "true" or "false".
.. warning:: If used incorrectly, the assumption that all tallies are spatially
separate can lead to incorrect results.
.. warning:: If used incorrectly, the assumption that all tallies are
spatially separate can lead to incorrect results.
*Default*: false
@ -1188,8 +1470,10 @@ element of the plots.xml is simply ``<plots>`` and any number output plots can
be defined with ``<plot>`` sub-elements. Two plot types are currently
implemented in openMC:
* ``slice`` 2D pixel plot along one of the major axes. Produces a PPM image file.
* ``voxel`` 3D voxel data dump. Produces a binary file containing voxel xyz position and cell or material id.
* ``slice`` 2D pixel plot along one of the major axes. Produces a PPM image
file.
* ``voxel`` 3D voxel data dump. Produces a binary file containing voxel xyz
position and cell or material id.
``<plot>`` Element
@ -1373,9 +1657,9 @@ attributes or sub-elements. These are not used in "voxel" plots:
CMFD Specification -- cmfd.xml
------------------------------
Coarse mesh finite difference acceleration method has been implemented in OpenMC.
Currently, it allows users to accelerate fission source convergence during
inactive neutron batches. To run CMFD, the ``<run_cmfd>`` element in
Coarse mesh finite difference acceleration method has been implemented in
OpenMC. Currently, it allows users to accelerate fission source convergence
during inactive neutron batches. To run CMFD, the ``<run_cmfd>`` element in
``settings.xml`` should be set to "true".
``<begin>`` Element
@ -1385,22 +1669,6 @@ The ``<begin>`` element controls what batch CMFD calculations should begin.
*Default*: 1
``<display>`` Element
---------------------
The ``<display>`` element sets one additional CMFD output column. Options are:
* "balance" - prints the RMS [%] of the resdiual from the neutron balance equation
on CMFD tallies.
* "dominance" - prints the estimated dominance ratio from the CMFD iterations.
**This will only work for power iteration eigensolver**.
* "entropy" - prints the *entropy* of the CMFD predicted fission source.
**Can only be used if OpenMC entropy is active as well**.
* "source" - prints the RMS [%] between the OpenMC fission source and CMFD
fission source.
*Default*: balance
``<dhat_reset>`` Element
------------------------
@ -1410,6 +1678,22 @@ It can be turned on with "true" and off with "false".
*Default*: false
``<display>`` Element
---------------------
The ``<display>`` element sets one additional CMFD output column. Options are:
* "balance" - prints the RMS [%] of the resdiual from the neutron balance
equation on CMFD tallies.
* "dominance" - prints the estimated dominance ratio from the CMFD iterations.
**This will only work for power iteration eigensolver**.
* "entropy" - prints the *entropy* of the CMFD predicted fission source.
**Can only be used if OpenMC entropy is active as well**.
* "source" - prints the RMS [%] between the OpenMC fission source and CMFD
fission source.
*Default*: balance
``<downscatter>`` Element
-------------------------
@ -1434,20 +1718,10 @@ It can be turned on with "true" and off with "false".
The ``<gauss_seidel_tolerance>`` element specifies two parameters. The first is
the absolute inner tolerance for Gauss-Seidel iterations when performing CMFD
and the second is the relative inner tolerance for Gauss-Seidel iterations
for CMFD calculations. It is only used in the standalone CMFD power iteration
solver and not when PETSc is active.
for CMFD calculations.
*Default*: 1.e-10 1.e-5
``<ksp_monitor>`` Element
-------------------------
The ``<ksp_monitor>`` element is used to view the convergence of linear GMRES
iterations in PETSc. This option can be turned on with "true" and turned off
with "false".
*Default*: false
``<ktol>`` Element
--------------------
@ -1463,11 +1737,11 @@ The CMFD mesh is a structured Cartesian mesh. This element has the following
attributes/sub-elements:
:lower_left:
The lower-left corner of the structured mesh. If only two coordinate are
The lower-left corner of the structured mesh. If only two coordinates are
given, it is assumed that the mesh is an x-y mesh.
:upper_right:
The upper-right corner of the structrued mesh. If only two coordinate are
The upper-right corner of the structrued mesh. If only two coordinates are
given, it is assumed that the mesh is an x-y mesh.
:dimension:
@ -1490,7 +1764,7 @@ attributes/sub-elements:
:map:
An optional acceleration map can be specified to overlay on the coarse
mesh spatial grid. If this option is used a ``1`` is used for a
mesh spatial grid. If this option is used, a ``1`` is used for a
non-accelerated region and a ``2`` is used for an accelerated region.
For a simple 4x4 coarse mesh with a 2x2 fuel lattice surrounded by
reflector, the map is:
@ -1527,8 +1801,8 @@ not impact the calculation.
``<power_monitor>`` Element
---------------------------
The ``<power_monitor>`` element is used to view the convergence of power iteration.
This option can be turned on with "true" and turned off with "false".
The ``<power_monitor>`` element is used to view the convergence of power
iteration. This option can be turned on with "true" and turned off with "false".
*Default*: false
@ -1536,26 +1810,16 @@ This option can be turned on with "true" and turned off with "false".
-------------------------
The ``<run_adjoint>`` element can be turned on with "true" to have an adjoint
calculation be performed on the last batch when CMFD is active. OpenMC should be
compiled with PETSc when using this option.
calculation be performed on the last batch when CMFD is active.
*Default*: false
``<solver>`` Element
--------------------
The ``<solver>`` element controls whether the CMFD eigenproblem is solved with
standard power iteration or nonlinear Jacobian-free Newton Krylov (JFNK).
By setting "power", power iteration is used and by setting "jfnk", JFNK is used.
*Default*: power
``<shift>`` Element
--------------------
The ``<shfit>`` element specifies an optional Wielandt shift parameter for
accelerating power iterations. It can only be used when PETSc is not active.
It is by default very large so the impact of the shift is effectively zero.
The ``<shift>`` element specifies an optional Wielandt shift parameter for
accelerating power iterations. It is by default very large so the impact of the
shift is effectively zero.
*Default*: 1e6
@ -1564,10 +1828,9 @@ It is by default very large so the impact of the shift is effectively zero.
The ``<spectral>`` element specifies an optional spectral radius that can be set to
accelerate the convergence of Gauss-Seidel iterations during CMFD power iteration
solve. Note this is only used in the standalone CMFD solver and does not affect
the calculation when PETSc is active.
solve.
*Default*: power
*Default*: 0.0
``<stol>`` Element
------------------
@ -1588,10 +1851,9 @@ should be reset.
``<write_matrices>`` Element
----------------------------
The ``<write_matrices>`` element is used to view the PETSc sparse matrices
created when solving CMFD equations. These binary output files can be imported
into MATLAB using PETSc-MATLAB utilities. This option can be
turned on with "true" and off with "false".
The ``<write_matrices>`` element is used to write the sparse matrices created
when solving CMFD equations. This option can be turned on with "true" and off
with "false".
*Default*: false

77
docs/source/usersguide/install.rst
View file

@ -86,21 +86,6 @@ Prerequisites
You may omit ``--enable-parallel`` if you want to compile HDF5_ in serial.
* PETSc_ for CMFD acceleration
To enable CMFD acceleration, you will need to have PETSc_ (3.4.2 or higher)
installed on your computer. The installed version will need to have been
compiled with the same compiler you intend to compile OpenMC with. OpenMC
requires PETSc_ to be configured with Fortran datatypes. An example of
configuring PETSc_ is listed below::
./configure --prefix=/opt/petsc/3.4.4 --download-f-blas-lapack \
--with-mpi-dir=/opt/mpich/3.1 --with-shared-libraries \
--with-fortran-datatypes
The BLAS/LAPACK library is not required to be downloaded and can be linked
explicitly (e.g., Intel MKL library).
* git_ version control software for obtaining source code
.. _gfortran: http://gcc.gnu.org/wiki/GFortran
@ -108,7 +93,6 @@ Prerequisites
.. _OpenMPI: http://www.open-mpi.org
.. _MPICH: http://www.mpich.org
.. _HDF5: http://www.hdfgroup.org/HDF5/
.. _PETSc: http://www.mcs.anl.gov/petsc/
Obtaining the Source
--------------------
@ -123,12 +107,12 @@ with GitHub since this involves setting up ssh_ keys. With git installed and
setup, the following command will download the full source code from the GitHub
repository::
git clone git://github.com/mit-crpg/openmc.git
git clone https://github.com/mit-crpg/openmc.git
By default, the cloned repository will be set to the development branch. To
switch to the source of the latest stable release, run the following commands::
cd openmc/src
cd openmc
git checkout master
.. _GitHub: https://github.com/mit-crpg/openmc
@ -139,23 +123,22 @@ Build Configuration
-------------------
Compiling OpenMC with CMake is carried out in two steps. First, ``cmake`` is run
to determine the compiler, whether optional packages (MPI, HDF5, PETSc) are
available, to generate a list of dependencies between source files so that they
may be compiled in the correct order, and to generate a normal Makefile. The
Makefile is then used by ``make`` to actually carry out the compile and linking
to determine the compiler, whether optional packages (MPI, HDF5) are available,
to generate a list of dependencies between source files so that they may be
compiled in the correct order, and to generate a normal Makefile. The Makefile
is then used by ``make`` to actually carry out the compile and linking
commands. A typical out-of-source build would thus look something like the
following
.. code-block:: sh
mkdir src/build
cd src/build
mkdir build && cd build
cmake ..
make
Note that first a build directory is created as a subdirectory of the source
directory. The Makefile in ``src/`` will automatically perform an out-of-source
build with default options.
directory. The Makefile in the top-level directory will automatically perform an
out-of-source build with default options.
CMakeLists.txt Options
++++++++++++++++++++++
@ -177,16 +160,21 @@ openmp
Enables shared-memory parallelism using the OpenMP API. The Fortran compiler
being used must support OpenMP.
petsc
Enables PETSc for use in CMFD acceleration. The PETSC_DIR variable should be
set to the base directory of the PETSc installation.
coverage
Compile and link code instrumented for coverage analysis. This is typically
used in conjunction with gcov_.
maxcoord
Maximum number of nested coordinate levels in geometry. Defaults to 10.
To set any of these options (e.g. turning on debug mode), the following form
should be used:
.. code-block:: sh
cmake -Ddebug=on /path/to/src
cmake -Ddebug=on /path/to/openmc
.. _gcov: https://gcc.gnu.org/onlinedocs/gcc/Gcov.html
Compiling with MPI
++++++++++++++++++
@ -197,14 +185,14 @@ the MPI Fortran wrapper. For example, in a bash shell:
.. code-block:: sh
export FC=mpif90
cmake /path/to/src
cmake /path/to/openmc
Note that in many shells, an environment variable can be set for a single
command, i.e.
.. code-block:: sh
FC=mpif90 cmake /path/to/src
FC=mpif90 cmake /path/to/openmc
Compiling with HDF5
+++++++++++++++++++
@ -215,14 +203,14 @@ the HDF5 Fortran wrapper. For example, in a bash shell:
.. code-block:: sh
export FC=h5fc
cmake /path/to/src
cmake /path/to/openmc
As noted above, an environment variable can typically be set for a single
command, i.e.
.. code-block:: sh
FC=h5fc cmake /path/to/src
FC=h5fc cmake /path/to/openmc
To compile with support for both MPI and HDF5, use the parallel HDF5 wrapper
``h5pfc`` instead. Note that this requires that your HDF5 installation be
@ -236,8 +224,7 @@ the root directory of the source code:
.. code-block:: sh
mkdir src/build
cd src/build
mkdir build && cd build
cmake ..
make
make install
@ -287,7 +274,8 @@ the source code root directory:
.. code-block:: sh
cd src
mkdir build && cd build
cmake ..
make
This will build an executable named ``openmc``.
@ -313,7 +301,6 @@ in the root directory of the OpenMC distribution:
.. code-block:: sh
cd src
make
This will build an executable named ``openmc``.
@ -332,7 +319,6 @@ the source directory and run the following:
.. code-block:: sh
cd src
make test
If you want more options for testing you can use ctest_ command. For example,
@ -340,7 +326,7 @@ if we wanted to run only the plot tests with 4 processors, we run:
.. code-block:: sh
cd src/build
cd build
ctest -j 4 -R plot
If you want to run the full test suite with different build options please
@ -385,12 +371,12 @@ the following steps must be taken:
2. In the root directory, a file named ``xsdir``, or some variant thereof,
should be present. This file contains a listing of all the cross sections and
is used by MCNP. This file should be converted to a ``cross_sections.xml``
file for use with OpenMC. A Python script is provided in the OpenMC
distribution for this purpose:
file for use with OpenMC. A utility is provided in the OpenMC distribution
for this purpose:
.. code-block:: sh
openmc/src/utils/convert_xsdir.py xsdir31 cross_sections.xml
openmc/scripts/openmc-xsdir-to-xml xsdir31 cross_sections.xml
3. In the converted ``cross_sections.xml`` file, change the contents of the
<directory> element to the absolute path of the directory containing the
@ -436,9 +422,8 @@ Running OpenMC
Once you have a model built (see :ref:`usersguide_input`), you can either run
the openmc executable directly from the directory containing your XML input
files, or you can specify as a command-line argument the directory containing
the XML input files. For example, if the path of your OpenMC executable is
``/home/username/openmc/src/openmc`` and your XML input files are in the
directory ``/home/username/somemodel/``, one way to run the simulation would be:
the XML input files. For example, if your XML input files are in the directory
``/home/username/somemodel/``, one way to run the simulation would be:
.. code-block:: sh

46
docs/source/usersguide/processing.rst
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@ -22,7 +22,7 @@ Most of these are easily obtainable in Ubuntu through the package manager, or
are easily installed with distutils.
.. [1] Required for tally data extraction from statepoints with statepoint.py
.. [2] Required only if reading HDF5 statepoint files.
.. [2] Required only if reading HDF5 statepoint files.
.. [3] Optional for plotting utilities
----------------------
@ -265,18 +265,18 @@ two heatmaps in the previous figure.
.. code-block:: python
#!/usr/bin/env python
import os
import statepoint
# load and parse the statepoint file
sp = statepoint.StatePoint('statepoint.300.binary')
sp.read_results()
tallyid = 0 # This is tally 1
score = 0 # This corresponds to flux (see tally.scores)
# get mesh dimensions
meshid = sp.tallies[tallyid].filters['mesh'].bins[0]
for i,m in enumerate(sp.meshes):
@ -299,7 +299,7 @@ two heatmaps in the previous figure.
[('mesh',(x,y,z)),('energyin',1)],
score)
fast[(x,y,z)] = val
# sum up the axial values and write datafile for gnuplot
with open('meshdata.dat','w') as fh:
for x in range(1,nx+1):
@ -336,7 +336,7 @@ Plotting in 3D
As with 3D plots of the geometry, meshtally data needs to be put into a standard
format for viewing. The utility statepoint_3d.py is provided to accomplish this
for both VTK and SILO. By default statepoint_3d.py processes a statepoint into a
3D file with all mesh tallies and filter/score combinations,
3D file with all mesh tallies and filter/score combinations,
.. code-block:: sh
@ -348,8 +348,8 @@ certain data arrays in order to keep file sizes down.
.. code-block:: sh
<openmc_root>/src/utils/statepoint_3d.py <statepoint_file> --tallies 2,4 --scores 4.1,4.3 -o output.silo
<openmc_root>/src/utils/statepoint_3d.py <statepoint_file> --filters 2.energyin.1 --vtk -o output.vtm
statepoint_3d.py <statepoint_file> --tallies 2,4 --scores 4.1,4.3 -o output.silo
statepoint_3d.py <statepoint_file> --filters 2.energyin.1 --vtk -o output.vtm
All available options for specifying a subset of tallies, scores, and filters
can be listed with the ``--list`` or ``-l`` command line options.
@ -358,7 +358,7 @@ can be listed with the ``--list`` or ``-l`` command line options.
VTK needs to use a multi-block dataset, which stores each mesh piece
in a different file in a subfolder. All meshes can be loaded at once
with the main VTM file, or each VTI file in the subfolder can be
loaded individually.
loaded individually.
Alternatively, the user can write their own Python script to manipulate the data
appropriately before insertion into a SILO or VTK file. For instance, if the
@ -396,7 +396,7 @@ and the equivalent VTK file with:
grid.SetOrigin(*mesh.lower_left)
grid.SetSpacing(*mesh.width)
# vtk cell arrays have x on the inners, so we need to reorder the data
# vtk cell arrays have x on the inners, so we need to reorder the data
idata = {}
for x in range(nx):
for y in range(ny):
@ -416,7 +416,7 @@ and the equivalent VTK file with:
grid.GetCellData().AddArray(vtkfastdata)
grid.GetCellData().AddArray(vtkthermaldata)
writer = vtk.vtkXMLImageDataWriter()
writer.SetInput(grid)
writer.SetFileName('tally.vti')
@ -486,16 +486,16 @@ example of an interactive ipython session using the statepoint.py Python module:
.. code-block:: python
In [1]: import statepoint
In [2]: sp = statepoint.StatePoint('statepoint.100.h5')
In [3]: sp.read_source()
In [4]: len(sp.source)
Out[4]: 1000
In [5]: sp.source[0:10]
Out[5]:
Out[5]:
[<SourceSite: xyz=[ 2.21980946 -8.92686048 87.93720485] at E=0.932923263566>,
<SourceSite: xyz=[ 2.21980946 -8.92686048 87.93720485] at E=0.349240220512>,
<SourceSite: xyz=[-31.21542213 -30.26762771 72.10845757] at E=3.75843584486>,
@ -506,17 +506,17 @@ example of an interactive ipython session using the statepoint.py Python module:
<SourceSite: xyz=[ -32.80427668 -15.49316628 125.26301151] at E=1.61907104162>,
<SourceSite: xyz=[ 53.20376026 -15.38643708 120.58071044] at E=3.33962024907>,
<SourceSite: xyz=[ 53.20376026 -15.38643708 120.58071044] at E=1.90185680329>]
In [6]: site = sp.source[0]
In [7]: site.weight
Out[7]: 1.0
In [8]: site.xyz
Out[8]: array([ 2.21980946, -8.92686048, 87.93720485])
In [9]: site.uvw
Out[9]: array([ 0.06740523, 0.50612814, 0.85982024])
In [10]: site.E
Out[10]: 0.93292326356564159

68
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@ -0,0 +1,68 @@
The file notebook_sphinxext.py was derived from code in PyNE and yt.
PyNE has the following license:
-------------------------------------------------------------------------------
Copyright 2011-2015, the PyNE Development Team. All rights reserved.
Redistribution and use in source and binary forms, with or without modification, are
permitted provided that the following conditions are met:
1. Redistributions of source code must retain the above copyright notice, this list of
conditions and the following disclaimer.
2. Redistributions in binary form must reproduce the above copyright notice, this list
of conditions and the following disclaimer in the documentation and/or other materials
provided with the distribution.
THIS SOFTWARE IS PROVIDED BY THE PYNE DEVELOPMENT TEAM ``AS IS'' AND ANY EXPRESS OR IMPLIED
WARRANTIES, INCLUDING, BUT NOT LIMITED TO, THE IMPLIED WARRANTIES OF MERCHANTABILITY AND
FITNESS FOR A PARTICULAR PURPOSE ARE DISCLAIMED. IN NO EVENT SHALL <COPYRIGHT HOLDER> OR
CONTRIBUTORS BE LIABLE FOR ANY DIRECT, INDIRECT, INCIDENTAL, SPECIAL, EXEMPLARY, OR
CONSEQUENTIAL DAMAGES (INCLUDING, BUT NOT LIMITED TO, PROCUREMENT OF SUBSTITUTE GOODS OR
SERVICES; LOSS OF USE, DATA, OR PROFITS; OR BUSINESS INTERRUPTION) HOWEVER CAUSED AND ON
ANY THEORY OF LIABILITY, WHETHER IN CONTRACT, STRICT LIABILITY, OR TORT (INCLUDING
NEGLIGENCE OR OTHERWISE) ARISING IN ANY WAY OUT OF THE USE OF THIS SOFTWARE, EVEN IF
ADVISED OF THE POSSIBILITY OF SUCH DAMAGE.
The views and conclusions contained in the software and documentation are those of the
authors and should not be interpreted as representing official policies, either expressed
or implied, of the stakeholders of the PyNE project or the employers of PyNE developers.
-------------------------------------------------------------------------------
yt has the following license:
-------------------------------------------------------------------------------
yt is licensed under the terms of the Modified BSD License (also known as New
or Revised BSD), as follows:
Copyright (c) 2013-, yt Development Team
Copyright (c) 2006-2013, Matthew Turk <matthewturk@gmail.com>
All rights reserved.
Redistribution and use in source and binary forms, with or without
modification, are permitted provided that the following conditions are met:
Redistributions of source code must retain the above copyright notice, this
list of conditions and the following disclaimer.
Redistributions in binary form must reproduce the above copyright notice, this
list of conditions and the following disclaimer in the documentation and/or
other materials provided with the distribution.
Neither the name of the yt Development Team nor the names of its
contributors may be used to endorse or promote products derived from this
software without specific prior written permission.
THIS SOFTWARE IS PROVIDED BY THE COPYRIGHT HOLDERS AND CONTRIBUTORS "AS IS" AND
ANY EXPRESS OR IMPLIED WARRANTIES, INCLUDING, BUT NOT LIMITED TO, THE IMPLIED
WARRANTIES OF MERCHANTABILITY AND FITNESS FOR A PARTICULAR PURPOSE ARE
DISCLAIMED. IN NO EVENT SHALL THE COPYRIGHT OWNER OR CONTRIBUTORS BE LIABLE
FOR ANY DIRECT, INDIRECT, INCIDENTAL, SPECIAL, EXEMPLARY, OR CONSEQUENTIAL
DAMAGES (INCLUDING, BUT NOT LIMITED TO, PROCUREMENT OF SUBSTITUTE GOODS OR
SERVICES; LOSS OF USE, DATA, OR PROFITS; OR BUSINESS INTERRUPTION) HOWEVER
CAUSED AND ON ANY THEORY OF LIABILITY, WHETHER IN CONTRACT, STRICT LIABILITY,
OR TORT (INCLUDING NEGLIGENCE OR OTHERWISE) ARISING IN ANY WAY OUT OF THE USE
OF THIS SOFTWARE, EVEN IF ADVISED OF THE POSSIBILITY OF SUCH DAMAGE.
-------------------------------------------------------------------------------

120
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@ -0,0 +1,120 @@
import sys
import os.path
import re
import time
from docutils import io, nodes, statemachine, utils
try:
from docutils.utils.error_reporting import ErrorString # the new way
except ImportError:
from docutils.error_reporting import ErrorString # the old way
from docutils.parsers.rst import Directive, convert_directive_function
from docutils.parsers.rst import directives, roles, states
from docutils.parsers.rst.roles import set_classes
from docutils.transforms import misc
try:
from IPython.nbconver.exporters import html
except ImportError:
from IPython.nbconvert import html
class Notebook(Directive):
"""Use nbconvert to insert a notebook into the environment.
This is based on the Raw directive in docutils
"""
required_arguments = 1
optional_arguments = 0
final_argument_whitespace = True
option_spec = {}
has_content = False
def run(self):
# check if raw html is supported
if not self.state.document.settings.raw_enabled:
raise self.warning('"%s" directive disabled.' % self.name)
# set up encoding
attributes = {'format': 'html'}
encoding = self.options.get(
'encoding', self.state.document.settings.input_encoding)
e_handler = self.state.document.settings.input_encoding_error_handler
# get path to notebook
source_dir = os.path.dirname(
os.path.abspath(self.state.document.current_source))
nb_path = os.path.normpath(os.path.join(source_dir,
self.arguments[0]))
nb_path = utils.relative_path(None, nb_path)
# convert notebook to html
exporter = html.HTMLExporter(template_file='full')
output, resources = exporter.from_filename(nb_path)
header = output.split('<head>', 1)[1].split('</head>',1)[0]
body = output.split('<body>', 1)[1].split('</body>',1)[0]
# add HTML5 scoped attribute to header style tags
header = header.replace('<style', '<style scoped="scoped"')
header = header.replace('body {\n overflow: visible;\n padding: 8px;\n}\n',
'')
header = header.replace("code,pre{", "code{")
# Filter out styles that conflict with the sphinx theme.
filter_strings = [
'navbar',
'body{',
'alert{',
'uneditable-input{',
'collapse{',
]
filter_strings.extend(['h%s{' % (i+1) for i in range(6)])
line_begin = [
'pre{',
'p{margin'
]
filterfunc = lambda x: not any([s in x for s in filter_strings])
header_lines = filter(filterfunc, header.split('\n'))
filterfunc = lambda x: not any([x.startswith(s) for s in line_begin])
header_lines = filter(filterfunc, header_lines)
header = '\n'.join(header_lines)
# concatenate raw html lines
lines = ['<div class="ipynotebook">']
lines.append(header)
lines.append(body)
lines.append('</div>')
text = '\n'.join(lines)
# add dependency
self.state.document.settings.record_dependencies.add(nb_path)
attributes['source'] = nb_path
# create notebook node
nb_node = notebook('', text, **attributes)
(nb_node.source, nb_node.line) = \
self.state_machine.get_source_and_line(self.lineno)
return [nb_node]
class notebook(nodes.raw):
pass
def visit_notebook_node(self, node):
self.visit_raw(node)
def depart_notebook_node(self, node):
self.depart_raw(node)
def setup(app):
app.add_node(notebook,
html=(visit_notebook_node, depart_notebook_node))
app.add_directive('notebook', Notebook)

141
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@ -0,0 +1,141 @@
import openmc
###############################################################################
# Simulation Input File Parameters
###############################################################################
# OpenMC simulation parameters
batches = 15
inactive = 5
particles = 10000
###############################################################################
# Exporting to OpenMC materials.xml File
###############################################################################
# Instantiate some Nuclides
h1 = openmc.Nuclide('H-1')
o16 = openmc.Nuclide('O-16')
u235 = openmc.Nuclide('U-235')
# Instantiate some Materials and register the appropriate Nuclides
moderator = openmc.Material(material_id=41, name='moderator')
moderator.set_density('g/cc', 1.0)
moderator.add_nuclide(h1, 2.)
moderator.add_nuclide(o16, 1.)
moderator.add_s_alpha_beta('HH2O', '71t')
fuel = openmc.Material(material_id=40, name='fuel')
fuel.set_density('g/cc', 4.5)
fuel.add_nuclide(u235, 1.)
# Instantiate a MaterialsFile, register all Materials, and export to XML
materials_file = openmc.MaterialsFile()
materials_file.default_xs = '71c'
materials_file.add_materials([moderator, fuel])
materials_file.export_to_xml()
###############################################################################
# Exporting to OpenMC geometry.xml File
###############################################################################
# Instantiate ZCylinder surfaces
surf1 = openmc.ZCylinder(surface_id=1, x0=0, y0=0, R=7, name='surf 1')
surf2 = openmc.ZCylinder(surface_id=2, x0=0, y0=0, R=9, name='surf 2')
surf3 = openmc.ZCylinder(surface_id=3, x0=0, y0=0, R=11, name='surf 3')
surf3.boundary_type = 'vacuum'
# Instantiate Cells
cell1 = openmc.Cell(cell_id=1, name='cell 1')
cell2 = openmc.Cell(cell_id=100, name='cell 2')
cell3 = openmc.Cell(cell_id=101, name='cell 3')
cell4 = openmc.Cell(cell_id=2, name='cell 4')
# Register Surfaces with Cells
cell1.add_surface(surface=surf2, halfspace=-1)
cell2.add_surface(surface=surf1, halfspace=-1)
cell3.add_surface(surface=surf1, halfspace=+1)
cell4.add_surface(surface=surf2, halfspace=+1)
cell4.add_surface(surface=surf3, halfspace=-1)
# Register Materials with Cells
cell2.fill = fuel
cell3.fill = moderator
cell4.fill = moderator
# Instantiate Universes
universe1 = openmc.Universe(universe_id=37)
root = openmc.Universe(universe_id=0, name='root universe')
cell1.fill = universe1
# Register Cells with Universes
universe1.add_cells([cell2, cell3])
root.add_cells([cell1, cell4])
# Instantiate a Geometry and register the root Universe
geometry = openmc.Geometry()
geometry.root_universe = root
# Instantiate a GeometryFile, register Geometry, and export to XML
geometry_file = openmc.GeometryFile()
geometry_file.geometry = geometry
geometry_file.export_to_xml()
###############################################################################
# Exporting to OpenMC settings.xml File
###############################################################################
# Instantiate a SettingsFile, set all runtime parameters, and export to XML
settings_file = openmc.SettingsFile()
settings_file.batches = batches
settings_file.inactive = inactive
settings_file.particles = particles
settings_file.set_source_space('box', [-4, -4, -4, 4, 4, 4])
settings_file.export_to_xml()
###############################################################################
# Exporting to OpenMC tallies.xml File
###############################################################################
# Instantiate some tally Filters
cell_filter = openmc.Filter(type='cell', bins=100)
energy_filter = openmc.Filter(type='energy', bins=[0., 20.])
energyout_filter = openmc.Filter(type='energyout', bins=[0., 20.])
# Instantiate the first Tally
first_tally = openmc.Tally(tally_id=1, name='first tally')
first_tally.add_filter(cell_filter)
scores = ['total', 'scatter', 'nu-scatter', \
'absorption', 'fission', 'nu-fission']
for score in scores:
first_tally.add_score(score)
# Instantiate the second Tally
second_tally = openmc.Tally(tally_id=2, name='second tally')
second_tally.add_filter(cell_filter)
second_tally.add_filter(energy_filter)
scores = ['total', 'scatter', 'nu-scatter', \
'absorption', 'fission', 'nu-fission']
for score in scores:
second_tally.add_score(score)
# Instantiate the third Tally
third_tally = openmc.Tally(tally_id=3, name='third tally')
third_tally.add_filter(cell_filter)
third_tally.add_filter(energy_filter)
third_tally.add_filter(energyout_filter)
scores = ['scatter', 'nu-scatter', 'nu-fission']
for score in scores:
third_tally.add_score(score)
# Instantiate a TalliesFile, register all Tallies, and export to XML
tallies_file = openmc.TalliesFile()
tallies_file.add_tally(first_tally)
tallies_file.add_tally(second_tally)
tallies_file.add_tally(third_tally)
tallies_file.export_to_xml()

176
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@ -0,0 +1,176 @@
import openmc
###############################################################################
# Simulation Input File Parameters
###############################################################################
# OpenMC simulation parameters
batches = 20
inactive = 10
particles = 10000
###############################################################################
# Exporting to OpenMC materials.xml File
###############################################################################
# Instantiate some Nuclides
h1 = openmc.Nuclide('H-1')
o16 = openmc.Nuclide('O-16')
u235 = openmc.Nuclide('U-235')
fe56 = openmc.Nuclide('Fe-56')
# Instantiate some Materials and register the appropriate Nuclides
fuel = openmc.Material(material_id=1, name='fuel')
fuel.set_density('g/cc', 4.5)
fuel.add_nuclide(u235, 1.)
moderator = openmc.Material(material_id=2, name='moderator')
moderator.set_density('g/cc', 1.0)
moderator.add_nuclide(h1, 2.)
moderator.add_nuclide(o16, 1.)
moderator.add_s_alpha_beta('HH2O', '71t')
iron = openmc.Material(material_id=3, name='iron')
iron.set_density('g/cc', 7.9)
iron.add_nuclide(fe56, 1.)
# Instantiate a MaterialsFile, register all Materials, and export to XML
materials_file = openmc.MaterialsFile()
materials_file.default_xs = '71c'
materials_file.add_materials([moderator, fuel, iron])
materials_file.export_to_xml()
###############################################################################
# Exporting to OpenMC geometry.xml File
###############################################################################
# Instantiate Surfaces
left = openmc.XPlane(surface_id=1, x0=-3, name='left')
right = openmc.XPlane(surface_id=2, x0=3, name='right')
bottom = openmc.YPlane(surface_id=3, y0=-4, name='bottom')
top = openmc.YPlane(surface_id=4, y0=4, name='top')
fuel_surf = openmc.ZCylinder(surface_id=5, x0=0, y0=0, R=0.4)
left.boundary_type = 'vacuum'
right.boundary_type = 'vacuum'
top.boundary_type = 'vacuum'
bottom.boundary_type = 'vacuum'
# Instantiate Cells
cell1 = openmc.Cell(cell_id=1, name='Cell 1')
cell2 = openmc.Cell(cell_id=101, name='cell 2')
cell3 = openmc.Cell(cell_id=102, name='cell 3')
cell4 = openmc.Cell(cell_id=500, name='cell 4')
cell5 = openmc.Cell(cell_id=600, name='cell 5')
cell6 = openmc.Cell(cell_id=601, name='cell 6')
# Register Surfaces with Cells
cell1.add_surface(left, halfspace=+1)
cell1.add_surface(right, halfspace=-1)
cell1.add_surface(bottom, halfspace=+1)
cell1.add_surface(top, halfspace=-1)
cell2.add_surface(fuel_surf, halfspace=-1)
cell3.add_surface(fuel_surf, halfspace=+1)
cell5.add_surface(fuel_surf, halfspace=-1)
cell6.add_surface(fuel_surf, halfspace=+1)
# Register Materials with Cells
cell2.fill = fuel
cell3.fill = moderator
cell4.fill = moderator
cell5.fill = iron
cell6.fill = moderator
# Instantiate Universe
univ1 = openmc.Universe(universe_id=1)
univ2 = openmc.Universe(universe_id=3)
univ3 = openmc.Universe(universe_id=4)
root = openmc.Universe(universe_id=0, name='root universe')
# Register Cells with Universe
univ1.add_cells([cell2, cell3])
univ2.add_cells([cell4])
univ3.add_cells([cell5, cell6])
root.add_cell(cell1)
# Instantiate a Lattice
lattice = openmc.HexLattice(lattice_id=5)
lattice.center = [0., 0., 0.]
lattice.pitch = [1., 2.]
lattice.universes = \
[ [ [univ2] + [univ3]*11, [univ2] + [univ3]*5, [univ3] ],
[ [univ2] + [univ1]*11, [univ2] + [univ1]*5, [univ1] ],
[ [univ2] + [univ3]*11, [univ2] + [univ3]*5, [univ3] ] ]
lattice.outer = univ2
# Fill Cell with the Lattice
cell1.fill = lattice
# Instantiate a Geometry and register the root Universe
geometry = openmc.Geometry()
geometry.root_universe = root
# Instantiate a GeometryFile, register Geometry, and export to XML
geometry_file = openmc.GeometryFile()
geometry_file.geometry = geometry
geometry_file.export_to_xml()
###############################################################################
# Exporting to OpenMC settings.xml File
###############################################################################
# Instantiate a SettingsFile, set all runtime parameters, and export to XML
settings_file = openmc.SettingsFile()
settings_file.batches = batches
settings_file.inactive = inactive
settings_file.particles = particles
settings_file.set_source_space('box', [-1, -1, -1, 1, 1, 1])
settings_file.keff_trigger = {'type' : 'std_dev', 'threshold' : 5E-4}
settings_file.trigger_active = True
settings_file.trigger_max_batches = 100
settings_file.export_to_xml()
###############################################################################
# Exporting to OpenMC plots.xml File
###############################################################################
plot_xy = openmc.Plot(plot_id=1)
plot_xy.filename = 'plot_xy'
plot_xy.origin = [0, 0, 0]
plot_xy.width = [6, 6]
plot_xy.pixels = [400, 400]
plot_xy.color = 'mat'
plot_yz = openmc.Plot(plot_id=2)
plot_yz.filename = 'plot_yz'
plot_yz.basis = 'yz'
plot_yz.origin = [0, 0, 0]
plot_yz.width = [8, 8]
plot_yz.pixels = [400, 400]
plot_yz.color = 'mat'
# Instantiate a PlotsFile, add Plot, and export to XML
plot_file = openmc.PlotsFile()
plot_file.add_plot(plot_xy)
plot_file.add_plot(plot_yz)
plot_file.export_to_xml()
###############################################################################
# Exporting to OpenMC tallies.xml File
###############################################################################
# Instantiate a distribcell Tally
tally = openmc.Tally(tally_id=1)
tally.add_filter(openmc.Filter(type='distribcell', bins=[cell2.id]))
tally.add_score('total')
# Instantiate a TalliesFile, register Tally/Mesh, and export to XML
tallies_file = openmc.TalliesFile()
tallies_file.add_tally(tally)
tallies_file.export_to_xml()

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import openmc
###############################################################################
# Simulation Input File Parameters
###############################################################################
# OpenMC simulation parameters
batches = 20
inactive = 10
particles = 10000
###############################################################################
# Exporting to OpenMC materials.xml File
###############################################################################
# Instantiate some Nuclides
h1 = openmc.Nuclide('H-1')
o16 = openmc.Nuclide('O-16')
u235 = openmc.Nuclide('U-235')
# Instantiate some Materials and register the appropriate Nuclides
fuel = openmc.Material(material_id=1, name='fuel')
fuel.set_density('g/cc', 4.5)
fuel.add_nuclide(u235, 1.)
moderator = openmc.Material(material_id=2, name='moderator')
moderator.set_density('g/cc', 1.0)
moderator.add_nuclide(h1, 2.)
moderator.add_nuclide(o16, 1.)
moderator.add_s_alpha_beta('HH2O', '71t')
# Instantiate a MaterialsFile, register all Materials, and export to XML
materials_file = openmc.MaterialsFile()
materials_file.default_xs = '71c'
materials_file.add_materials([moderator, fuel])
materials_file.export_to_xml()
###############################################################################
# Exporting to OpenMC geometry.xml File
###############################################################################
# Instantiate Surfaces
left = openmc.XPlane(surface_id=1, x0=-2, name='left')
right = openmc.XPlane(surface_id=2, x0=2, name='right')
bottom = openmc.YPlane(surface_id=3, y0=-2, name='bottom')
top = openmc.YPlane(surface_id=4, y0=2, name='top')
fuel1 = openmc.ZCylinder(surface_id=5, x0=0, y0=0, R=0.4)
fuel2 = openmc.ZCylinder(surface_id=6, x0=0, y0=0, R=0.3)
fuel3 = openmc.ZCylinder(surface_id=7, x0=0, y0=0, R=0.2)
left.boundary_type = 'vacuum'
right.boundary_type = 'vacuum'
top.boundary_type = 'vacuum'
bottom.boundary_type = 'vacuum'
# Instantiate Cells
cell1 = openmc.Cell(cell_id=1, name='Cell 1')
cell2 = openmc.Cell(cell_id=2, name='Cell 2')
cell3 = openmc.Cell(cell_id=101, name='cell 3')
cell4 = openmc.Cell(cell_id=102, name='cell 4')
cell5 = openmc.Cell(cell_id=201, name='cell 5')
cell6 = openmc.Cell(cell_id=202, name='cell 6')
cell7 = openmc.Cell(cell_id=301, name='cell 7')
cell8 = openmc.Cell(cell_id=302, name='cell 8')
# Register Surfaces with Cells
cell1.add_surface(left, halfspace=+1)
cell1.add_surface(right, halfspace=-1)
cell1.add_surface(bottom, halfspace=+1)
cell1.add_surface(top, halfspace=-1)
cell2.add_surface(left, halfspace=+1)
cell2.add_surface(right, halfspace=-1)
cell2.add_surface(bottom, halfspace=+1)
cell2.add_surface(top, halfspace=-1)
cell3.add_surface(fuel1, halfspace=-1)
cell4.add_surface(fuel1, halfspace=+1)
cell5.add_surface(fuel2, halfspace=-1)
cell6.add_surface(fuel2, halfspace=+1)
cell7.add_surface(fuel3, halfspace=-1)
cell8.add_surface(fuel3, halfspace=+1)
# Register Materials with Cells
cell3.fill = fuel
cell4.fill = moderator
cell5.fill = fuel
cell6.fill = moderator
cell7.fill = fuel
cell8.fill = moderator
# Instantiate Universe
univ1 = openmc.Universe(universe_id=1)
univ2 = openmc.Universe(universe_id=2)
univ3 = openmc.Universe(universe_id=3)
univ4 = openmc.Universe(universe_id=5)
root = openmc.Universe(universe_id=0, name='root universe')
# Register Cells with Universe
univ1.add_cells([cell3, cell4])
univ2.add_cells([cell5, cell6])
univ3.add_cells([cell7, cell8])
root.add_cell(cell1)
univ4.add_cell(cell2)
# Instantiate nested Lattices
lattice1 = openmc.RectLattice(lattice_id=4, name='4x4 assembly')
lattice1.dimension = [2, 2]
lattice1.lower_left = [-1., -1.]
lattice1.pitch = [1., 1.]
lattice1.universes = [[univ1, univ2],
[univ2, univ3]]
lattice2 = openmc.RectLattice(lattice_id=6, name='4x4 core')
lattice2.dimension = [2, 2]
lattice2.lower_left = [-2., -2.]
lattice2.pitch = [2., 2.]
lattice2.universes = [[univ4, univ4],
[univ4, univ4]]
# Fill Cell with the Lattice
cell1.fill = lattice2
cell2.fill = lattice1
# Instantiate a Geometry and register the root Universe
geometry = openmc.Geometry()
geometry.root_universe = root
# Instantiate a GeometryFile, register Geometry, and export to XML
geometry_file = openmc.GeometryFile()
geometry_file.geometry = geometry
geometry_file.export_to_xml()
###############################################################################
# Exporting to OpenMC settings.xml File
###############################################################################
# Instantiate a SettingsFile, set all runtime parameters, and export to XML
settings_file = openmc.SettingsFile()
settings_file.batches = batches
settings_file.inactive = inactive
settings_file.particles = particles
settings_file.set_source_space('box', [-1, -1, -1, 1, 1, 1])
settings_file.export_to_xml()
###############################################################################
# Exporting to OpenMC plots.xml File
###############################################################################
plot = openmc.Plot(plot_id=1)
plot.origin = [0, 0, 0]
plot.width = [4, 4]
plot.pixels = [400, 400]
plot.color = 'mat'
# Instantiate a PlotsFile, add Plot, and export to XML
plot_file = openmc.PlotsFile()
plot_file.add_plot(plot)
plot_file.export_to_xml()
###############################################################################
# Exporting to OpenMC tallies.xml File
###############################################################################
# Instantiate a tally mesh
mesh = openmc.Mesh(mesh_id=1)
mesh.type = 'rectangular'
mesh.dimension = [4, 4]
mesh.lower_left = [-2, -2]
mesh.width = [1, 1]
# Instantiate tally Filter
mesh_filter = openmc.Filter()
mesh_filter.mesh = mesh
# Instantiate the Tally
tally = openmc.Tally(tally_id=1)
tally.add_filter(mesh_filter)
tally.add_score('total')
# Instantiate a TalliesFile, register Tally/Mesh, and export to XML
tallies_file = openmc.TalliesFile()
tallies_file.add_mesh(mesh)
tallies_file.add_tally(tally)
tallies_file.export_to_xml()

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import openmc
###############################################################################
# Simulation Input File Parameters
###############################################################################
# OpenMC simulation parameters
batches = 20
inactive = 10
particles = 10000
###############################################################################
# Exporting to OpenMC materials.xml File
###############################################################################
# Instantiate some Nuclides
h1 = openmc.Nuclide('H-1')
o16 = openmc.Nuclide('O-16')
u235 = openmc.Nuclide('U-235')
# Instantiate some Materials and register the appropriate Nuclides
fuel = openmc.Material(material_id=1, name='fuel')
fuel.set_density('g/cc', 4.5)
fuel.add_nuclide(u235, 1.)
moderator = openmc.Material(material_id=2, name='moderator')
moderator.set_density('g/cc', 1.0)
moderator.add_nuclide(h1, 2.)
moderator.add_nuclide(o16, 1.)
moderator.add_s_alpha_beta('HH2O', '71t')
# Instantiate a MaterialsFile, register all Materials, and export to XML
materials_file = openmc.MaterialsFile()
materials_file.default_xs = '71c'
materials_file.add_materials([moderator, fuel])
materials_file.export_to_xml()
###############################################################################
# Exporting to OpenMC geometry.xml File
###############################################################################
# Instantiate Surfaces
left = openmc.XPlane(surface_id=1, x0=-2, name='left')
right = openmc.XPlane(surface_id=2, x0=2, name='right')
bottom = openmc.YPlane(surface_id=3, y0=-2, name='bottom')
top = openmc.YPlane(surface_id=4, y0=2, name='top')
fuel1 = openmc.ZCylinder(surface_id=5, x0=0, y0=0, R=0.4)
fuel2 = openmc.ZCylinder(surface_id=6, x0=0, y0=0, R=0.3)
fuel3 = openmc.ZCylinder(surface_id=7, x0=0, y0=0, R=0.2)
left.boundary_type = 'vacuum'
right.boundary_type = 'vacuum'
top.boundary_type = 'vacuum'
bottom.boundary_type = 'vacuum'
# Instantiate Cells
cell1 = openmc.Cell(cell_id=1, name='Cell 1')
cell2 = openmc.Cell(cell_id=101, name='cell 2')
cell3 = openmc.Cell(cell_id=102, name='cell 3')
cell4 = openmc.Cell(cell_id=201, name='cell 4')
cell5 = openmc.Cell(cell_id=202, name='cell 5')
cell6 = openmc.Cell(cell_id=301, name='cell 6')
cell7 = openmc.Cell(cell_id=302, name='cell 7')
# Register Surfaces with Cells
cell1.add_surface(left, halfspace=+1)
cell1.add_surface(right, halfspace=-1)
cell1.add_surface(bottom, halfspace=+1)
cell1.add_surface(top, halfspace=-1)
cell2.add_surface(fuel1, halfspace=-1)
cell3.add_surface(fuel1, halfspace=+1)
cell4.add_surface(fuel2, halfspace=-1)
cell5.add_surface(fuel2, halfspace=+1)
cell6.add_surface(fuel3, halfspace=-1)
cell7.add_surface(fuel3, halfspace=+1)
# Register Materials with Cells
cell2.fill = fuel
cell3.fill = moderator
cell4.fill = fuel
cell5.fill = moderator
cell6.fill = fuel
cell7.fill = moderator
# Instantiate Universe
univ1 = openmc.Universe(universe_id=1)
univ2 = openmc.Universe(universe_id=2)
univ3 = openmc.Universe(universe_id=3)
root = openmc.Universe(universe_id=0, name='root universe')
# Register Cells with Universe
univ1.add_cells([cell2, cell3])
univ2.add_cells([cell4, cell5])
univ3.add_cells([cell6, cell7])
root.add_cell(cell1)
# Instantiate a Lattice
lattice = openmc.RectLattice(lattice_id=5)
lattice.dimension = [4, 4]
lattice.lower_left = [-2., -2.]
lattice.pitch = [1., 1.]
lattice.universes = [[univ1, univ2, univ1, univ2],
[univ2, univ3, univ2, univ3],
[univ1, univ2, univ1, univ2],
[univ2, univ3, univ2, univ3]]
# Fill Cell with the Lattice
cell1.fill = lattice
# Instantiate a Geometry and register the root Universe
geometry = openmc.Geometry()
geometry.root_universe = root
# Instantiate a GeometryFile, register Geometry, and export to XML
geometry_file = openmc.GeometryFile()
geometry_file.geometry = geometry
geometry_file.export_to_xml()
###############################################################################
# Exporting to OpenMC settings.xml File
###############################################################################
# Instantiate a SettingsFile, set all runtime parameters, and export to XML
settings_file = openmc.SettingsFile()
settings_file.batches = batches
settings_file.inactive = inactive
settings_file.particles = particles
settings_file.set_source_space('box', [-1, -1, -1, 1, 1, 1])
settings_file.trigger_active = True
settings_file.trigger_max_batches = 100
settings_file.export_to_xml()
###############################################################################
# Exporting to OpenMC plots.xml File
###############################################################################
plot = openmc.Plot(plot_id=1)
plot.origin = [0, 0, 0]
plot.width = [4, 4]
plot.pixels = [400, 400]
plot.color = 'mat'
# Instantiate a PlotsFile, add Plot, and export to XML
plot_file = openmc.PlotsFile()
plot_file.add_plot(plot)
plot_file.export_to_xml()
###############################################################################
# Exporting to OpenMC tallies.xml File
###############################################################################
# Instantiate a tally mesh
mesh = openmc.Mesh(mesh_id=1)
mesh.type = 'rectangular'
mesh.dimension = [4, 4]
mesh.lower_left = [-2, -2]
mesh.width = [1, 1]
# Instantiate tally Filter
mesh_filter = openmc.Filter()
mesh_filter.mesh = mesh
# Instantiate tally Trigger
trigger = openmc.Trigger(trigger_type='rel_err', threshold=1E-2)
trigger.add_score('all')
# Instantiate the Tally
tally = openmc.Tally(tally_id=1)
tally.add_filter(mesh_filter)
tally.add_score('total')
tally.add_trigger(trigger)
# Instantiate a TalliesFile, register Tally/Mesh, and export to XML
tallies_file = openmc.TalliesFile()
tallies_file.add_mesh(mesh)
tallies_file.add_tally(tally)
tallies_file.export_to_xml()

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import openmc
###############################################################################
# Simulation Input File Parameters
###############################################################################
# OpenMC simulation parameters
batches = 100
inactive = 10
particles = 1000
###############################################################################
# Exporting to OpenMC materials.xml File
###############################################################################
# Instantiate some Nuclides
h1 = openmc.Nuclide('H-1')
h2 = openmc.Nuclide('H-2')
he4 = openmc.Nuclide('He-4')
b10 = openmc.Nuclide('B-10')
b11 = openmc.Nuclide('B-11')
o16 = openmc.Nuclide('O-16')
o17 = openmc.Nuclide('O-17')
cr50 = openmc.Nuclide('Cr-50')
cr52 = openmc.Nuclide('Cr-52')
cr53 = openmc.Nuclide('Cr-53')
cr54 = openmc.Nuclide('Cr-54')
fe54 = openmc.Nuclide('Fe-54')
fe56 = openmc.Nuclide('Fe-56')
fe57 = openmc.Nuclide('Fe-57')
fe58 = openmc.Nuclide('Fe-58')
zr90 = openmc.Nuclide('Zr-90')
zr91 = openmc.Nuclide('Zr-91')
zr92 = openmc.Nuclide('Zr-92')
zr94 = openmc.Nuclide('Zr-94')
zr96 = openmc.Nuclide('Zr-96')
sn112 = openmc.Nuclide('Sn-112')
sn114 = openmc.Nuclide('Sn-114')
sn115 = openmc.Nuclide('Sn-115')
sn116 = openmc.Nuclide('Sn-116')
sn117 = openmc.Nuclide('Sn-117')
sn118 = openmc.Nuclide('Sn-118')
sn119 = openmc.Nuclide('Sn-119')
sn120 = openmc.Nuclide('Sn-120')
sn122 = openmc.Nuclide('Sn-122')
sn124 = openmc.Nuclide('Sn-124')
u234 = openmc.Nuclide('U-234')
u235 = openmc.Nuclide('U-235')
u238 = openmc.Nuclide('U-238')
# Instantiate some Materials and register the appropriate Nuclides
uo2 = openmc.Material(material_id=1, name='UO2 fuel at 2.4% wt enrichment')
uo2.set_density('g/cm3', 10.29769)
uo2.add_nuclide(u234, 4.4843e-6)
uo2.add_nuclide(u235, 5.5815e-4)
uo2.add_nuclide(u238, 2.2408e-2)
uo2.add_nuclide(o16, 4.5829e-2)
uo2.add_nuclide(o17, 1.1164e-4)
helium = openmc.Material(material_id=2, name='Helium for gap')
helium.set_density('g/cm3', 0.001598)
helium.add_nuclide(he4, 2.4044e-4)
zircaloy = openmc.Material(material_id=3, name='Zircaloy 4')
zircaloy.set_density('g/cm3', 6.55)
zircaloy.add_nuclide(o16, 3.0743e-4)
zircaloy.add_nuclide(o17, 7.4887e-7)
zircaloy.add_nuclide(cr50, 3.2962e-6)
zircaloy.add_nuclide(cr52, 6.3564e-5)
zircaloy.add_nuclide(cr53, 7.2076e-6)
zircaloy.add_nuclide(cr54, 1.7941e-6)
zircaloy.add_nuclide(fe54, 8.6699e-6)
zircaloy.add_nuclide(fe56, 1.3610e-4)
zircaloy.add_nuclide(fe57, 3.1431e-6)
zircaloy.add_nuclide(fe58, 4.1829e-7)
zircaloy.add_nuclide(zr90, 2.1827e-2)
zircaloy.add_nuclide(zr91, 4.7600e-3)
zircaloy.add_nuclide(zr92, 7.2758e-3)
zircaloy.add_nuclide(zr94, 7.3734e-3)
zircaloy.add_nuclide(zr96, 1.1879e-3)
zircaloy.add_nuclide(sn112, 4.6735e-6)
zircaloy.add_nuclide(sn114, 3.1799e-6)
zircaloy.add_nuclide(sn115, 1.6381e-6)
zircaloy.add_nuclide(sn116, 7.0055e-5)
zircaloy.add_nuclide(sn117, 3.7003e-5)
zircaloy.add_nuclide(sn118, 1.1669e-4)
zircaloy.add_nuclide(sn119, 4.1387e-5)
zircaloy.add_nuclide(sn120, 1.5697e-4)
zircaloy.add_nuclide(sn122, 2.2308e-5)
zircaloy.add_nuclide(sn124, 2.7897e-5)
borated_water = openmc.Material(material_id=4, name='Borated water at 975 ppm')
borated_water.set_density('g/cm3', 0.740582)
borated_water.add_nuclide(b10, 8.0042e-6)
borated_water.add_nuclide(b11, 3.2218e-5)
borated_water.add_nuclide(h1, 4.9457e-2)
borated_water.add_nuclide(h2, 7.4196e-6)
borated_water.add_nuclide(o16, 2.4672e-2)
borated_water.add_nuclide(o17, 6.0099e-5)
borated_water.add_s_alpha_beta('HH2O', '71t')
# Instantiate a MaterialsFile, register all Materials, and export to XML
materials_file = openmc.MaterialsFile()
materials_file.default_xs = '71c'
materials_file.add_materials([uo2, helium, zircaloy, borated_water])
materials_file.export_to_xml()
###############################################################################
# Exporting to OpenMC geometry.xml File
###############################################################################
# Instantiate ZCylinder surfaces
fuel_or = openmc.ZCylinder(surface_id=1, x0=0, y0=0, R=0.39218, name='Fuel OR')
clad_ir = openmc.ZCylinder(surface_id=2, x0=0, y0=0, R=0.40005, name='Clad IR')
clad_or = openmc.ZCylinder(surface_id=3, x0=0, y0=0, R=0.45720, name='Clad OR')
left = openmc.XPlane(surface_id=4, x0=-0.62992, name='left')
right = openmc.XPlane(surface_id=5, x0=0.62992, name='right')
bottom = openmc.YPlane(surface_id=6, y0=-0.62992, name='bottom')
top = openmc.YPlane(surface_id=7, y0=0.62992, name='top')
left.boundary_type = 'reflective'
right.boundary_type = 'reflective'
top.boundary_type = 'reflective'
bottom.boundary_type = 'reflective'
# Instantiate Cells
fuel = openmc.Cell(cell_id=1, name='cell 1')
gap = openmc.Cell(cell_id=2, name='cell 2')
clad = openmc.Cell(cell_id=3, name='cell 3')
water = openmc.Cell(cell_id=4, name='cell 4')
# Register Surfaces with Cells
fuel.add_surface(fuel_or, halfspace=-1)
gap.add_surface(fuel_or, halfspace=+1)
gap.add_surface(clad_ir, halfspace=-1)
clad.add_surface(clad_ir, halfspace=+1)
clad.add_surface(clad_or, halfspace=-1)
water.add_surface(clad_or, halfspace=+1)
water.add_surface(left, halfspace=+1)
water.add_surface(right, halfspace=-1)
water.add_surface(bottom, halfspace=+1)
water.add_surface(top, halfspace=-1)
# Register Materials with Cells
fuel.fill = uo2
gap.fill = helium
clad.fill = zircaloy
water.fill = borated_water
# Instantiate Universe
root = openmc.Universe(universe_id=0, name='root universe')
# Register Cells with Universe
root.add_cells([fuel, gap, clad, water])
# Instantiate a Geometry and register the root Universe
geometry = openmc.Geometry()
geometry.root_universe = root
# Instantiate a GeometryFile, register Geometry, and export to XML
geometry_file = openmc.GeometryFile()
geometry_file.geometry = geometry
geometry_file.export_to_xml()
###############################################################################
# Exporting to OpenMC settings.xml File
###############################################################################
# Instantiate a SettingsFile, set all runtime parameters, and export to XML
settings_file = openmc.SettingsFile()
settings_file.batches = batches
settings_file.inactive = inactive
settings_file.particles = particles
settings_file.set_source_space('box', [-0.62992, -0.62992, -1, \
0.62992, 0.62992, 1])
settings_file.entropy_lower_left = [-0.39218, -0.39218, -1.e50]
settings_file.entropy_upper_right = [0.39218, 0.39218, 1.e50]
settings_file.entropy_dimension = [10, 10, 1]
settings_file.export_to_xml()
###############################################################################
# Exporting to OpenMC tallies.xml File
###############################################################################
# Instantiate a tally mesh
mesh = openmc.Mesh(mesh_id=1)
mesh.type = 'rectangular'
mesh.dimension = [100, 100, 1]
mesh.lower_left = [-0.62992, -0.62992, -1.e50]
mesh.upper_right = [0.62992, 0.62992, 1.e50]
# Instantiate some tally Filters
energy_filter = openmc.Filter(type='energy', bins=[0., 4.e-6, 20.])
mesh_filter = openmc.Filter()
mesh_filter.mesh = mesh
# Instantiate the Tally
tally = openmc.Tally(tally_id=1, name='tally 1')
tally.add_filter(energy_filter)
tally.add_filter(mesh_filter)
tally.add_score('flux')
tally.add_score('fission')
tally.add_score('nu-fission')
# Instantiate a TalliesFile, register all Tallies, and export to XML
tallies_file = openmc.TalliesFile()
tallies_file.add_mesh(mesh)
tallies_file.add_tally(tally)
tallies_file.export_to_xml()

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import openmc
###############################################################################
# Simulation Input File Parameters
###############################################################################
# OpenMC simulation parameters
batches = 500
inactive = 10
particles = 10000
###############################################################################
# Exporting to OpenMC materials.xml File
###############################################################################
# Instantiate a Nuclides
u235 = openmc.Nuclide('U-235')
# Instantiate a Material and register the Nuclide
fuel = openmc.Material(material_id=1, name='fuel')
fuel.set_density('g/cc', 4.5)
fuel.add_nuclide(u235, 1.)
# Instantiate a MaterialsFile, register Material, and export to XML
materials_file = openmc.MaterialsFile()
materials_file.default_xs = '71c'
materials_file.add_material(fuel)
materials_file.export_to_xml()
###############################################################################
# Exporting to OpenMC geometry.xml File
###############################################################################
# Instantiate Surfaces
surf1 = openmc.XPlane(surface_id=1, x0=-1, name='surf 1')
surf2 = openmc.XPlane(surface_id=2, x0=+1, name='surf 2')
surf3 = openmc.YPlane(surface_id=3, y0=-1, name='surf 3')
surf4 = openmc.YPlane(surface_id=4, y0=+1, name='surf 4')
surf5 = openmc.ZPlane(surface_id=5, z0=-1, name='surf 5')
surf6 = openmc.ZPlane(surface_id=6, z0=+1, name='surf 6')
surf1.boundary_type = 'vacuum'
surf2.boundary_type = 'vacuum'
surf3.boundary_type = 'reflective'
surf4.boundary_type = 'reflective'
surf5.boundary_type = 'reflective'
surf6.boundary_type = 'reflective'
# Instantiate Cell
cell = openmc.Cell(cell_id=1, name='cell 1')
# Register Surfaces with Cell
cell.add_surface(surface=surf1, halfspace=+1)
cell.add_surface(surface=surf2, halfspace=-1)
cell.add_surface(surface=surf3, halfspace=+1)
cell.add_surface(surface=surf4, halfspace=-1)
cell.add_surface(surface=surf5, halfspace=+1)
cell.add_surface(surface=surf6, halfspace=-1)
# Register Material with Cell
cell.fill = fuel
# Instantiate Universes
root = openmc.Universe(universe_id=0, name='root universe')
# Register Cell with Universe
root.add_cell(cell)
# Instantiate a Geometry and register the root Universe
geometry = openmc.Geometry()
geometry.root_universe = root
# Instantiate a GeometryFile, register Geometry, and export to XML
geometry_file = openmc.GeometryFile()
geometry_file.geometry = geometry
geometry_file.export_to_xml()
###############################################################################
# Exporting to OpenMC settings.xml File
###############################################################################
# Instantiate a SettingsFile, set all runtime parameters, and export to XML
settings_file = openmc.SettingsFile()
settings_file.batches = batches
settings_file.inactive = inactive
settings_file.particles = particles
settings_file.set_source_space('box', [-1, -1, -1, 1, 1, 1])
settings_file.export_to_xml()

0
examples/basic/geometry.xml → examples/xml/basic/geometry.xml
View file

0
examples/basic/materials.xml → examples/xml/basic/materials.xml
View file

0
examples/basic/settings.xml → examples/xml/basic/settings.xml
View file

0
examples/basic/tallies.xml → examples/xml/basic/tallies.xml
View file

6
examples/lattice/nested/geometry.xml → examples/xml/lattice/nested/geometry.xml
View file

@ -12,10 +12,9 @@
<!-- 4 x 4 assembly -->
<lattice id="4">
<type>rectangular</type>
<dimension>2 2</dimension>
<lower_left>-1.0 -1.0</lower_left>
<width>1.0 1.0</width>
<pitch>1.0 1.0</pitch>
<universes>
1 2
2 3
@ -24,10 +23,9 @@
<!-- 4 x 4 core -->
<lattice id="6">
<type>rectangular</type>
<dimension>2 2</dimension>
<lower_left>-2.0 -2.0</lower_left>
<width>2.0 2.0</width>
<pitch>2.0 2.0</pitch>
<universes>
5 5
5 5

0
examples/lattice/nested/materials.xml → examples/xml/lattice/nested/materials.xml
View file

0
examples/lattice/nested/plots.xml → examples/xml/lattice/nested/plots.xml
View file

0
examples/lattice/nested/settings.xml → examples/xml/lattice/nested/settings.xml
View file

0
examples/lattice/nested/tallies.xml → examples/xml/lattice/nested/tallies.xml
View file

3
examples/lattice/simple/geometry.xml → examples/xml/lattice/simple/geometry.xml
View file

@ -10,10 +10,9 @@
<cell id="302" universe="3" material="2" surfaces="7" />
<lattice id="5">
<type>rectangular</type>
<dimension>4 4</dimension>
<lower_left>-2.0 -2.0</lower_left>
<width>1.0 1.0</width>
<pitch>1.0 1.0</pitch>
<universes>
1 2 1 2
2 3 2 3

0
examples/lattice/simple/materials.xml → examples/xml/lattice/simple/materials.xml
View file

0
examples/lattice/simple/plots.xml → examples/xml/lattice/simple/plots.xml
View file

0
examples/lattice/simple/settings.xml → examples/xml/lattice/simple/settings.xml
View file

0
examples/lattice/simple/tallies.xml → examples/xml/lattice/simple/tallies.xml
View file

0
examples/pincell/geometry.xml → examples/xml/pincell/geometry.xml
View file

0
examples/pincell/materials.xml → examples/xml/pincell/materials.xml
View file

0
examples/pincell/settings.xml → examples/xml/pincell/settings.xml
View file

0
examples/pincell/tallies.xml → examples/xml/pincell/tallies.xml
View file

0
examples/reflective/geometry.xml → examples/xml/reflective/geometry.xml
View file

0
examples/reflective/materials.xml → examples/xml/reflective/materials.xml
View file

0
examples/reflective/settings.xml → examples/xml/reflective/settings.xml
View file

19
openmc/__init__.py Normal file
View file

@ -0,0 +1,19 @@
from openmc.element import *
from openmc.geometry import *
from openmc.nuclide import *
from openmc.material import *
from openmc.plots import *
from openmc.settings import *
from openmc.surface import *
from openmc.universe import *
from openmc.mesh import *
from openmc.filter import *
from openmc.trigger import *
from openmc.tallies import *
from openmc.cmfd import *
from openmc.executor import *
try:
from openmc.opencg_compatible import *
except ImportError:
pass

12
src/utils/convert_binary.py → openmc/ace.py Executable file → Normal file
View file

@ -1,9 +1,5 @@
#!/usr/bin/env python
from __future__ import division
from struct import pack
import sys
def ascii_to_binary(ascii_file, binary_file):
@ -67,11 +63,3 @@ def ascii_to_binary(ascii_file, binary_file):
# Close binary file
binary.close()
if __name__ == '__main__':
# Check for proper number of arguments
if len(sys.argv) < 3:
sys.exit('Usage: {0} ascii_file binary_file'.format(sys.argv[0]))
# Convert ASCII file
ascii_to_binary(sys.argv[1], sys.argv[2])

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