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2
docs/source/usersguide/beginners.rst
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@ -65,7 +65,7 @@ Now let's look at the pros and cons of Monte Carlo methods:
- **Pro**: Running simulations in parallel is conceptually very simple.
- **Con**: Because they related on repeated random sampling, they are
- **Con**: Because they rely on repeated random sampling, they are
computationally very expensive.
- **Con**: A simulation doesn't automatically give you the global solution

203
docs/source/usersguide/input.rst
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@ -281,6 +281,8 @@ based on the recommended value in LA-UR-14-24530_.
.. note:: This element is not used in the multi-group :ref:`energy_mode`.
.. _multipole_library:
``<multipole_library>`` Element
-------------------------------
@ -290,8 +292,8 @@ OpenMC can use it for on-the-fly Doppler-broadening of resolved resonance range
cross sections. If this element is absent from the settings.xml file, the
:envvar:`OPENMC_MULTIPOLE_LIBRARY` environment variable will be used.
.. note:: The <use_windowed_multipole> element must also be set to "true"
for windowed multipole functionality.
.. note:: The :ref:`temperature_method` must also be set to "multipole" for
windowed multipole functionality.
``<max_order>`` Element
---------------------------
@ -395,19 +397,16 @@ attributes or sub-elements:
:scatterer:
An element with attributes/sub-elements called ``nuclide``, ``method``,
``xs_label``, ``xs_label_0K``, ``E_min``, and ``E_max``. The ``nuclide``
attribute is the name, as given by the ``name`` attribute within the
``nuclide`` sub-element of the ``material`` element in ``materials.xml``,
of the nuclide to which a resonance scattering treatment is to be applied.
``E_min``, and ``E_max``. The ``nuclide`` attribute is the name, as given
by the ``name`` attribute within the ``nuclide`` sub-element of the
``material`` element in ``materials.xml``, of the nuclide to which a
resonance scattering treatment is to be applied.
The ``method`` attribute gives the type of resonance scattering treatment
that is to be applied to the ``nuclide``. Acceptable inputs - none of
which are case-sensitive - for the ``method`` attribute are ``ARES``,
``CXS``, ``WCM``, and ``DBRC``. Descriptions of each of these methods
are documented here_. The ``xs_label`` attribute gives the label for the
cross section data of the ``nuclide`` at a given temperature. The
``xs_label_0K`` gives the label for the 0 K cross section data for the
``nuclide``. The ``E_min`` attribute gives the minimum energy above
which the ``method`` is applied. The ``E_max`` attribute gives the
are documented here_. The ``E_min`` attribute gives the minimum energy
above which the ``method`` is applied. The ``E_max`` attribute gives the
maximum energy below which the ``method`` is applied. One example would
be as follows:
@ -419,16 +418,12 @@ attributes or sub-elements:
<scatterer>
<nuclide>U-238</nuclide>
<method>ARES</method>
<xs_label>92238.72c</xs_label>
<xs_label_0K>92238.00c</xs_label_0K>
<E_min>5.0e-6</E_min>
<E_max>40.0e-6</E_max>
</scatterer>
<scatterer>
<nuclide>Pu-239</nuclide>
<method>dbrc</method>
<xs_label>94239.72c</xs_label>
<xs_label_0K>94239.00c</xs_label_0K>
<E_min>0.01e-6</E_min>
<E_max>210.0e-6</E_max>
</scatterer>
@ -688,7 +683,7 @@ attributes/sub-elements:
*Default*: false
:source_write:
:write:
If this element is set to "false", source sites are not written
to the state point or source point file. This can substantially reduce the
size of state points if large numbers of particles per batch are used.
@ -714,6 +709,48 @@ survival biasing, otherwise known as implicit capture or absorption.
*Default*: false
.. _temperature_default:
``<temperature_default>`` Element
---------------------------------
The ``<temperature_default>`` element specifies a default temperature in Kelvin
that is to be applied to cells in the absence of an explicit cell temperature or
a material default temperature.
*Default*: 293.6 K
.. _temperature_method:
``<temperature_method>`` Element
--------------------------------
The ``<temperature_method>`` element has an accepted value of "nearest",
"interpolation", or "multipole". A value of "nearest" indicates that for each
cell, the nearest temperature at which cross sections are given is to be
applied, within a given tolerance (see :ref:`temperature_tolerance`). A value of
"interpolation" indicates that cross sections are to be linear-linear
interpolated between temperatures at which nuclear data are present (see
:ref:`temperature_treatment`). A value of "multipole" indicates that the
windowed multipole method should be used to evaluate temperature-dependent cross
sections in the resolved resonance range (a :ref:`windowed multipole library
<multipole_library>` must also be available).
*Default*: "nearest"
.. _temperature_tolerance:
``<temperature_tolerance>`` Element
-----------------------------------
The ``<temperature_tolerance>`` element specifies a tolerance in Kelvin that is
to be applied when the "nearest" temperature method is used. For example, if a
cell temperature is 340 K and the tolerance is 15 K, then the closest
temperature in the range of 325 K to 355 K will be used to evaluate cross
sections.
*Default*: 10 K
``<threads>`` Element
---------------------
@ -836,6 +873,35 @@ displayed. This element takes the following attributes:
*Default*: 5
``<volume_calc>`` Element
-------------------------
The ``<volume_calc>`` element indicates that a stochastic volume calculation
should be run at the beginning of the simulation. This element has the following
sub-elements/attributes:
:cells:
The unique IDs of cells for which the volume should be estimated.
*Default*: None
:samples:
The number of samples used to estimate volumes.
*Default*: None
:lower_left:
The lower-left Cartesian coordinates of a bounding box that is used to
sample points within.
*Default*: None
:upper_right:
The upper-right Cartesian coordinates of a bounding box that is used to
sample points within.
*Default*: None
--------------------------------------
Geometry Specification -- geometry.xml
--------------------------------------
@ -921,11 +987,19 @@ Each ``<surface>`` element can have the following attributes or sub-elements:
*Default*: None
:boundary:
The boundary condition for the surface. This can be "transmission",
"vacuum", or "reflective".
The boundary condition for the surface. This can be "transmission",
"vacuum", "reflective", or "periodic". Periodic boundary conditions can
only be applied to x-, y-, and z-planes. Only axis-aligned periodicity is
supported, i.e., x-planes can only be paired with x-planes. Specify which
planes are periodic and the code will automatically identify which planes
are paired together.
*Default*: "transmission"
:periodic_surface_id:
If a periodic boundary condition is applied, this attribute identifies the
``id`` of the corresponding periodic sufrace.
The following quadratic surfaces can be modeled:
:x-plane:
@ -1053,7 +1127,9 @@ Each ``<cell>`` element can have the following attributes or sub-elements:
specified for the "distributed temperature" feature. This will give each
unique instance of the cell its own temperature.
*Default*: The temperature of the coldest nuclide in the cell's material(s)
*Default*: If a material default temperature is supplied, it is used. In the
absence of a material default temperature, the :ref:`global default
temperature <temperature_default>` is used.
:rotation:
If the cell is filled with a universe, this element specifies the angles in
@ -1258,6 +1334,14 @@ Each ``material`` element can have the following attributes or sub-elements:
*Default*: ""
:temperature:
An element with no attributes which is used to set the default temperature
of the material in Kelvin.
*Default*: If a material default temperature is not given and a cell
temperature is not specified, the :ref:`global default temperature
<temperature_default>` is used.
:density:
An element with attributes/sub-elements called ``value`` and ``units``. The
``value`` attribute is the numeric value of the density while the ``units``
@ -1278,17 +1362,16 @@ Each ``material`` element can have the following attributes or sub-elements:
``nuclide``, ``element``, or ``sab`` quantity.
:nuclide:
An element with attributes/sub-elements called ``name``, ``xs``, and ``ao``
An element with attributes/sub-elements called ``name``, and ``ao``
or ``wo``. The ``name`` attribute is the name of the cross-section for a
desired nuclide while the ``xs`` attribute is the cross-section
identifier. Finally, the ``ao`` and ``wo`` attributes specify the atom or
desired nuclide. Finally, the ``ao`` and ``wo`` attributes specify the atom or
weight percent of that nuclide within the material, respectively. One
example would be as follows:
.. code-block:: xml
<nuclide name="H-1" xs="70c" ao="2.0" />
<nuclide name="O-16" xs="70c" ao="1.0" />
<nuclide name="H1" ao="2.0" />
<nuclide name="O16" ao="1.0" />
.. note:: If one nuclide is specified in atom percent, all others must also
be given in atom percent. The same applies for weight percentages.
@ -1312,11 +1395,10 @@ Each ``material`` element can have the following attributes or sub-elements:
Specifies that a natural element is present in the material. The natural
element is split up into individual isotopes based on `IUPAC Isotopic
Compositions of the Elements 2009`_. This element has
attributes/sub-elements called ``name``, ``xs``, and ``ao``. The ``name``
attribute is the atomic symbol of the element while the ``xs`` attribute is
the cross-section identifier. Finally, the ``ao`` attribute specifies the
atom percent of the element within the material, respectively. One example
would be as follows:
attributes/sub-elements called ``name``, and ``ao``. The ``name``
attribute is the atomic symbol of the element. Finally, the ``ao``
attribute specifies the atom percent of the element within the material,
respectively. One example would be as follows:
.. code-block:: xml
@ -1346,10 +1428,9 @@ Each ``material`` element can have the following attributes or sub-elements:
multi-group :ref:`energy_mode`.
:sab:
Associates an S(a,b) table with the material. This element has
attributes/sub-elements called ``name`` and ``xs``. The ``name`` attribute
is the name of the S(a,b) table that should be associated with the material,
and ``xs`` is the cross-section identifier for the table.
Associates an S(a,b) table with the material. This element has one
attribute/sub-element called ``name``. The ``name`` attribute
is the name of the S(a,b) table that should be associated with the material.
*Default*: None
@ -1360,14 +1441,13 @@ Each ``material`` element can have the following attributes or sub-elements:
recognizes that some multi-group libraries may be providing material
specific macroscopic cross sections instead of always providing nuclide
specific data like in the continuous-energy case. To that end, the
macroscopic element has attributes/sub-elements called ``name``, and ``xs``.
macroscopic element has one attribute/sub-element called ``name``.
The ``name`` attribute is the name of the cross-section for a
desired nuclide while the ``xs`` attribute is the cross-section
identifier. One example would be as follows:
desired nuclide. One example would be as follows:
.. code-block:: xml
<macroscopic name="UO2" xs="71c" />
<macroscopic name="UO2" />
.. note:: This element is only used in the multi-group :ref:`energy_mode`.
@ -1376,18 +1456,6 @@ Each ``material`` element can have the following attributes or sub-elements:
.. _IUPAC Isotopic Compositions of the Elements 2009:
http://pac.iupac.org/publications/pac/pdf/2011/pdf/8302x0397.pdf
``<default_xs>`` Element
------------------------
In some circumstances, the cross-section identifier may be the same for many or
all nuclides in a given problem. In this case, rather than specifying the
``xs=...`` attribute on every nuclide, a ``<default_xs>`` element can be used to
set the default cross-section identifier for any nuclide without an identifier
explicitly listed. This element has no attributes and accepts a 3-letter string
that indicates the default cross-section identifier, e.g. "70c".
*Default*: None
------------------------------------
Tallies Specification -- tallies.xml
------------------------------------
@ -1623,7 +1691,8 @@ The ``<tally>`` element accepts the following sub-elements:
|Score | Description |
+======================+===================================================+
|absorption |Total absorption rate. This accounts for all |
| |reactions which do not produce secondary neutrons. |
| |reactions which do not produce secondary neutrons |
| |as well as fission. |
+----------------------+---------------------------------------------------+
|elastic |Elastic scattering reaction rate. |
+----------------------+---------------------------------------------------+
@ -1755,6 +1824,10 @@ The ``<tally>`` element accepts the following sub-elements:
| |fission. This score type is not used in the |
| |multi-group :ref:`energy_mode`. |
+----------------------+---------------------------------------------------+
|prompt-nu-fission |Total production of prompt neutrons due to |
| |fission. This score type is not used in the |
| |multi-group :ref:`energy_mode`. |
+----------------------+---------------------------------------------------+
|nu-fission |Total production of neutrons due to fission. |
+----------------------+---------------------------------------------------+
|nu-scatter, |These scores are similar in functionality to their |
@ -1796,6 +1869,32 @@ The ``<tally>`` element accepts the following sub-elements:
| |:math:`\gamma`-rays are assumed to deposit their |
| |energy locally. Units are MeV per source particle. |
+----------------------+---------------------------------------------------+
|fission-q-prompt |The prompt fission energy production rate. This |
| |energy comes in the form of fission fragment |
| |nuclei, prompt neutrons, and prompt |
| |:math:`\gamma`-rays. This value depends on the |
| |incident energy and it requires that the nuclear |
| |data library contains the optional fission energy |
| |release data. Energy is assumed to be deposited |
| |locally. Units are MeV per source particle. |
+----------------------+---------------------------------------------------+
|fission-q-recoverable |The recoverable fission energy production rate. |
| |This energy comes in the form of fission fragment |
| |nuclei, prompt and delayed neutrons, prompt and |
| |delayed :math:`\gamma`-rays, and delayed |
| |:math:`\beta`-rays. This tally differs from the |
| |kappa-fission tally in that it is dependent on |
| |incident neutron energy and it requires that the |
| |nuclear data library contains the optional fission |
| |energy release data. Energy is assumed to be |
| |deposited locally. Units are MeV per source |
| |paticle. |
+----------------------+---------------------------------------------------+
|decay-rate |The delayed-nu-fission-weighted decay rate where |
| |the decay rate is in units of inverse seconds. |
| |This score type is not used in the |
| |multi-group :ref:`energy_mode`. |
+----------------------+---------------------------------------------------+
.. note::
The ``analog`` estimator is actually identical to the ``collision``
@ -2077,8 +2176,8 @@ attributes or sub-elements. These are not used in "voxel" plots:
*Default*: None
:meshlines:
The ``meshlines`` sub-element allows for plotting the boundaries of
a tally mesh on top of a plot. Only one ``meshlines`` element is allowed per
The ``meshlines`` sub-element allows for plotting the boundaries of a
regular mesh on top of a plot. Only one ``meshlines`` element is allowed per
``plot`` element, and it must contain as attributes or sub-elements a mesh
type and a linewidth. Optionally, a color may be specified for the overlay:

158
docs/source/usersguide/install.rst
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@ -204,20 +204,22 @@ should be used:
Compiling with MPI
++++++++++++++++++
To compile with MPI, set the :envvar:`FC` environment variable to the path to
the MPI Fortran wrapper. For example, in a bash shell:
To compile with MPI, set the :envvar:`FC` and :envvar:`CC` environment variables
to the path to the MPI Fortran and C wrappers, respectively. For example, in a
bash shell:
.. code-block:: sh
export FC=mpif90
export CC=mpicc
cmake /path/to/openmc
Note that in many shells, an environment variable can be set for a single
command, i.e.
Note that in many shells, environment variables can be set for a single command,
i.e.
.. code-block:: sh
FC=mpif90 cmake /path/to/openmc
FC=mpif90 CC=mpicc cmake /path/to/openmc
Selecting HDF5 Installation
+++++++++++++++++++++++++++
@ -343,7 +345,7 @@ compiler, it is necessary to specify that all objects be compiled with the
.. code-block:: sh
mkdir build && cd build
FC=ifort FFLAGS=-mmic cmake -Dopenmp=on ..
FC=ifort CC=icc FFLAGS=-mmic cmake -Dopenmp=on ..
make
Note that unless an HDF5 build for the Intel Xeon Phi is already on your target
@ -381,14 +383,16 @@ Cross Section Configuration
---------------------------
In order to run a simulation with OpenMC, you will need cross section data for
each nuclide or material in your problem. OpenMC can be run in
continuous-energy or multi-group mode.
each nuclide or material in your problem. OpenMC can be run in continuous-energy
or multi-group mode.
In continuous-energy mode OpenMC uses ACE format cross sections; in this case
you can use nuclear data that was processed with NJOY_, such as that
distributed with MCNP_ or Serpent_. Several sources provide free processed
ACE data as described below. The TALYS-based evaluated nuclear data library,
TENDL_, is also openly available in ACE format.
In continuous-energy mode, OpenMC uses a native HDF5 format to store all nuclear
data. If you have ACE format data that was produced with NJOY_, such as that
distributed with MCNP_ or Serpent_, it can be converted to the HDF5 format using
the :ref:`openmc-ace-to-hdf5 <other_cross_sections>` script distributed with
OpenMC. Several sources provide openly available ACE data as described
below. The TALYS-based evaluated nuclear data library, TENDL_, is also available
in ACE format.
In multi-group mode, OpenMC utilizes an XML-based library format which can be
used to describe nuclide- or material-specific quantities.
@ -398,8 +402,8 @@ Using ENDF/B-VII.1 Cross Sections from NNDC
The NNDC_ provides ACE data from the ENDF/B-VII.1 neutron and thermal scattering
sublibraries at four temperatures processed using NJOY_. To use this data with
OpenMC, a script is provided with OpenMC that will automatically download,
extract, and set up a confiuration file:
OpenMC, a script is provided with OpenMC that will automatically download and
extract the ACE data, fix any deficiencies, and create an HDF5 library:
.. code-block:: sh
@ -408,56 +412,99 @@ extract, and set up a confiuration file:
At this point, you should set the :envvar:`OPENMC_CROSS_SECTIONS` environment
variable to the absolute path of the file
``openmc/data/nndc/cross_sections.xml``. This cross section set is used by the
test suite.
``openmc/data/nndc_hdf5/cross_sections.xml``. This cross section set is used by
the test suite.
Using JEFF Cross Sections from OECD/NEA
---------------------------------------
The NEA_ provides processed ACE data from the JEFF_ nuclear library upon
request. A DVD of the data can be requested here_. To use this data with OpenMC,
the following steps must be taken:
The NEA_ provides processed ACE data from the JEFF_ library. To use this data
with OpenMC, a script is provided with OpenMC that will automatically download
and extract the ACE data, fix any deficiencies, and create an HDF5 library.
1. Copy and unzip the data on the DVD to a directory on your computer.
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 utility is provided in the OpenMC distribution
for this purpose:
.. code-block:: sh
.. code-block:: sh
cd openmc/data
python get_jeff_data.py
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
actual ACE files.
4. Additionally, you may need to change any occurrences of upper-case "ACE"
within the ``cross_sections.xml`` file to lower-case.
5. Either set the :ref:`cross_sections` in a settings.xml file or the
:envvar:`OPENMC_CROSS_SECTIONS` environment variable to the absolute path of
the ``cross_sections.xml`` file.
At this point, you should set the :envvar:`OPENMC_CROSS_SECTIONS` environment
variable to the absolute path of the file
``openmc/data/jeff-3.2-hdf5/cross_sections.xml``.
Using Cross Sections from MCNP
------------------------------
To use cross sections distributed with MCNP, change the <directory> element in
the ``cross_sections.xml`` file in the root directory of the OpenMC distribution
to the location of the MCNP cross sections. Then, either set the
:ref:`cross_sections` in a settings.xml file or the
:envvar:`OPENMC_CROSS_SECTIONS` environment variable to the absolute path of
the ``cross_sections.xml`` file.
OpenMC is provided with a script that will automatically convert ENDF/B-VII.0
and ENDF/B-VII.1 ACE data that is provided with MCNP5 or MCNP6. To convert the
ENDF/B-VII.0 ACE files (``endf70[a-k]`` and ``endf70sab``) into the native HDF5
format, run the following:
Using Cross Sections from Serpent
---------------------------------
.. code-block:: sh
cd openmc/data
python convert_mcnp_endf70.py /path/to/mcnpdata/
where ``/path/to/mcnpdata`` is the directory containing the ``endf70[a-k]``
files.
To convert the ENDF/B-VII.1 ACE files (the endf71x and ENDF71SaB libraries), use
the following script:
.. code-block:: sh
cd openmc/data
python convert_mcnp_endf71.py /path/to/mcnpdata
where ``/path/to/mcnpdata`` is the directory containing the ``endf71x`` and
``ENDF71SaB`` directories.
.. _other_cross_sections:
Using Other Cross Sections
--------------------------
If you have a library of ACE format cross sections other than those listed above
that you need to convert to OpenMC's HDF5 format, the ``openmc-ace-to-hdf5``
script can be used. There are four different ways you can specify ACE libraries
that are to be converted:
1. List each ACE library as a positional argument. This is very useful in
conjunction with the usual shell utilities (ls, find, etc.).
2. Use the --xml option to specify a pre-v0.9 cross_sections.xml file.
3. Use the --xsdir option to specify a MCNP xsdir file.
4. Use the --xsdata option to specify a Serpent xsdata file.
The script does not use any extra information from cross_sections.xml/ xsdir/
xsdata files to determine whether the nuclide is metastable. Instead, the
--metastable argument can be used to specify whether the ZAID naming convention
follows the NNDC data convention (1000*Z + A + 300 + 100*m), or the MCNP data
convention (essentially the same as NNDC, except that the first metastable state
of Am242 is 95242 and the ground state is 95642).
The ``openmc-ace-to-hdf5`` script has the following command-line flags:
-h, --help show this help message and exit
-d DESTINATION, --destination DESTINATION
Directory to create new library in (default: .)
-m META, --metastable META
How to interpret ZAIDs for metastable nuclides. META
can be either 'nndc' or 'mcnp'. (default: nndc)
--xml XML Old-style cross_sections.xml that lists ACE libraries
(default: None)
--xsdir XSDIR MCNP xsdir file that lists ACE libraries (default:
None)
--xsdata XSDATA Serpent xsdata file that lists ACE libraries (default:
None)
--fission_energy_release FISSION_ENERGY_RELEASE
HDF5 file containing fission energy release data
(default: None)
To use cross sections distributed with Serpent, change the <directory> element
in the ``cross_sections_serpent.xml`` file in the root directory of the OpenMC
distribution to the location of the Serpent cross sections. Then, either set the
:ref:`cross_sections` in a settings.xml file or the
:envvar:`OPENMC_CROSS_SECTIONS` environment variable to the absolute path of
the ``cross_sections_serpent.xml``
file.
Using Multi-Group Cross Sections
--------------------------------
@ -469,14 +516,13 @@ However, if the user has obtained or generated their own library, the user
should set the :envvar:`OPENMC_MG_CROSS_SECTIONS` environment variable
to the absolute path of the file library expected to used most frequently.
.. _NJOY: http://t2.lanl.gov/nis/codes.shtml
.. _NJOY: http://t2.lanl.gov/nis/codes/NJOY12/
.. _NNDC: http://www.nndc.bnl.gov/endf/b7.1/acefiles.html
.. _NEA: http://www.oecd-nea.org
.. _JEFF: http://www.oecd-nea.org/dbdata/jeff/
.. _here: http://www.oecd-nea.org/dbdata/pubs/jeff312-cd.html
.. _JEFF: https://www.oecd-nea.org/dbforms/data/eva/evatapes/jeff_32/
.. _MCNP: http://mcnp.lanl.gov
.. _Serpent: http://montecarlo.vtt.fi
.. _TENDL: ftp://ftp.nrg.eu/pub/www/talys/tendl2012/tendl2012.html
.. _TENDL: https://tendl.web.psi.ch/tendl_2015/tendl2015.html
--------------
Running OpenMC