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2
docs/source/usersguide/beginners.rst
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@ -12,7 +12,7 @@ In a nutshell, OpenMC simulates neutrons moving around randomly in a `nuclear
reactor`_ (or other fissile system). This is what's known as `Monte Carlo`_
simulation. Neutrons are important in nuclear reactors because they are the
particles that induce `fission`_ in uranium and other nuclides. Knowing the
behavior of neutrons allows you to figure out how often and where fission
behavior of neutrons allows you to determine how often and where fission
occurs. The amount of energy released is then directly proportional to the
fission reaction rate since most heat is produced by fission. By simulating many
neutrons (millions or billions), it is possible to determine the average

1
docs/source/usersguide/index.rst
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@ -14,6 +14,7 @@ essential aspects of using OpenMC to perform simulations.
beginners
install
input
mgxs_library
output/index
processing
troubleshoot

113
docs/source/usersguide/input.rst
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@ -112,9 +112,11 @@ standard deviation.
The ``<cross_sections>`` element has no attributes and simply indicates the path
to an XML cross section listing file (usually named cross_sections.xml). If this
element is absent from the settings.xml file, the :envvar:`CROSS_SECTIONS`
environment variable will be used to find the path to the XML cross section
listing.
element is absent from the settings.xml file, the
:envvar:`OPENMC_CROSS_SECTIONS` environment variable will be used to find the
path to the XML cross section listing when in continuous-energy mode, and the
:envvar:`OPENMC_MG_CROSS_SECTIONS` environment variable will be used in
multi-group mode.
``<cutoff>`` Element
--------------------
@ -212,8 +214,21 @@ cross section values between.
*Default*: logarithm
.. note:: This element is not used in the multi-group :ref:`energy_mode`.
.. _LA-UR-14-24530: https://laws.lanl.gov/vhosts/mcnp.lanl.gov/pdf_files/la-ur-14-24530.pdf
.. _energy_mode:
``<energy_mode>`` Element
-------------------------
The ``<energy_mode>`` element tells OpenMC if the run-mode should be
continuous-energy or multi-group. Options for entry are: ``continuous-energy``
or ``multi-group``.
*Default*: continuous-energy
``<entropy>`` Element
---------------------
@ -264,6 +279,8 @@ based on the recommended value in LA-UR-14-24530_.
*Default*: 8000
.. note:: This element is not used in the multi-group :ref:`energy_mode`.
``<multipole_library>`` Element
-------------------------------
@ -271,11 +288,24 @@ The ``<multipole_library>`` element indicates the directory containing a
windowed multipole library. If a windowed multipole library is available,
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:`MULTIPOLE_LIBRARY` environment variable will be used.
:envvar:`OPENMC_MULTIPOLE_LIBRARY` environment variable will be used.
.. note:: The <use_windowed_multipole> element must also be set to "True"
.. note:: The <use_windowed_multipole> element must also be set to "true"
for windowed multipole functionality.
``<max_order>`` Element
---------------------------
The ``<max_order>`` element allows the user to set a maximum scattering order
to apply to every nuclide/material in the problem. That is, if the data
library has :math:`P_3` data available, but ``<max_order>`` was set to ``1``,
then, OpenMC will only use up to the :math:`P_1` data.
*Default*: Use the maximum order in the data library
.. note:: This element is not used in the continuous-energy
:ref:`energy_mode`.
.. _natural_elements:
``<natural_elements>`` Element
@ -324,10 +354,10 @@ out the file and "false" will not.
*Default*: false
:summary:
Writes out an ASCII summary file describing all of the user input files that
Writes out an HDF5 summary file describing all of the user input files that
were read in.
*Default*: false
*Default*: true
:tallies:
Write out an ASCII file of tally results.
@ -355,6 +385,8 @@ or sub-elements and can be set to either "false" or "true".
*Default*: true
.. note:: This element is not used in the multi-group :ref:`energy_mode`.
``<resonance_scattering>`` Element
----------------------------------
@ -414,6 +446,8 @@ attributes or sub-elements:
*Defaults*: None (scatterer), ARES (method), 0.01 eV (E_min), 1.0 keV (E_max)
.. note:: This element is not used in the multi-group :ref:`energy_mode`.
``<run_cmfd>`` Element
----------------------
@ -591,6 +625,8 @@ variable and whose sub-elements/attributes are as follows:
number :math:`a` that parameterizes the distribution :math:`p(x) dx = c x
e^{-x/a} dx`.
.. note:: The above format should be used even when using the multi-group
:ref:`energy_mode`.
:interpolation:
For a "tabular" distribution, ``interpolation`` can be set to "histogram" or
"linear-linear" thereby specifying how tabular points are to be interpolated.
@ -1216,10 +1252,18 @@ Each ``material`` element can have the following attributes or sub-elements:
``<element>`` sub-elements are to be interpreted as nuclide/element
densities in atom/b-cm, and the total density of the material is taken as
the sum of all nuclides/elements. The "sum" option cannot be used in
conjunction with weight percents.
conjunction with weight percents. The "macro" unit is used with
a ``macroscopic`` quantity to indicate that the density is already included
in the library and thus not needed here. However, if a value is provided
for the ``value``, then this is treated as a number density multiplier on
the macroscopic cross sections in the multi-group data. This can be used,
for example, when perturbing the density slightly.
*Default*: None
.. note:: A ``macroscopic`` quantity can not be used in conjunction with a
``nuclide``, ``element``, or ``sab`` quantity.
:nuclide:
An element with attributes/sub-elements called ``name``, ``xs``, and ``ao``
or ``wo``. The ``name`` attribute is the name of the cross-section for a
@ -1247,6 +1291,9 @@ Each ``material`` element can have the following attributes or sub-elements:
*Default*: None
.. note:: The ``scattering`` attribute/sub-element is not used in the
multi-group :ref:`energy_mode`.
:element:
Specifies that a natural element is present in the material. The natural
@ -1282,6 +1329,9 @@ Each ``material`` element can have the following attributes or sub-elements:
*Default*: None
.. note:: The ``scattering`` attribute/sub-element is not used in the
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
@ -1290,6 +1340,26 @@ Each ``material`` element can have the following attributes or sub-elements:
*Default*: None
.. note:: This element is not used in the multi-group :ref:`energy_mode`.
:macroscopic:
The ``macroscopic`` element is similar to the ``nuclide`` element, but,
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``.
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:
.. code-block:: xml
<macroscopic name="UO2" xs="71c" />
.. note:: This element is only used in the multi-group :ref:`energy_mode`.
*Default*: None
.. _IUPAC Isotopic Compositions of the Elements 2009:
http://pac.iupac.org/publications/pac/pdf/2011/pdf/8302x0397.pdf
@ -1376,7 +1446,8 @@ The ``<tally>`` element accepts the following sub-elements:
A list of universes for which the tally should be accumulated.
:energy:
A monotonically increasing list of bounding **pre-collision** energies
In continuous-energy mode, this filter should be provided as a
monotonically increasing list of bounding **pre-collision** energies
for a number of groups. For example, if this filter is specified as
.. code-block:: xml
@ -1386,17 +1457,24 @@ The ``<tally>`` element accepts the following sub-elements:
then two energy bins will be created, one with energies between 0 and
1 MeV and the other with energies between 1 and 20 MeV.
In multi-group mode the bins provided must match group edges
defined in the multi-group library.
:energyout:
A monotonically increasing list of bounding **post-collision**
energies for a number of groups. For example, if this filter is
specified as
In continuous-energy mode, this filter should be provided as a
monotonically increasing list of bounding **post-collision** energies
for a number of groups. For example, if this filter is specified as
.. code-block:: xml
<filter type="energyout" bins="0.0 1.0 20.0" />
then two post-collision energy bins will be created, one with energies
between 0 and 1 MeV and the other with energies between 1 and 20 MeV.
then two post-collision energy bins will be created, one with
energies between 0 and 1 MeV and the other with energies between
1 and 20 MeV.
In multi-group mode the bins provided must match group edges
defined in the multi-group library.
:mu:
A monotonically increasing list of bounding **post-collision** cosines
@ -1480,6 +1558,8 @@ The ``<tally>`` element accepts the following sub-elements:
<filter type="delayedgroup" bins="1 2 3 4 5 6" />
.. note:: This filter type is not used in the multi-group :ref:`energy_mode`.
: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
@ -1659,7 +1739,8 @@ The ``<tally>`` element accepts the following sub-elements:
|Score | Description |
+======================+===================================================+
|delayed-nu-fission |Total production of delayed neutrons due to |
| |fission. |
| |fission. This score type is not used in the |
| |multi-group :ref:`energy_mode`. |
+----------------------+---------------------------------------------------+
|nu-fission |Total production of neutrons due to fission. |
+----------------------+---------------------------------------------------+
@ -1688,6 +1769,8 @@ The ``<tally>`` element accepts the following sub-elements:
+----------------------+---------------------------------------------------+
|inverse-velocity |The flux-weighted inverse velocity where the |
| |velocity is in units of centimeters per second. |
| |This score type is not used in the |
| |multi-group :ref:`energy_mode`. |
+----------------------+---------------------------------------------------+
|kappa-fission |The recoverable energy production rate due to |
| |fission. The recoverable energy is defined as the |

50
docs/source/usersguide/install.rst
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@ -227,6 +227,7 @@ your PATH environment variable and subsequently uses it to determine library
locations and compile flags. If you have multiple installations of HDF5 or one
that does not appear on your PATH, you can set the HDF5_ROOT environment
variable to the root directory of the HDF5 installation, e.g.
.. code-block:: sh
export HDF5_ROOT=/opt/hdf5/1.8.15
@ -355,7 +356,7 @@ Testing Build
-------------
If you have ENDF/B-VII.1 cross sections from NNDC_ you can test your build.
Make sure the **CROSS_SECTIONS** environmental variable is set to the
Make sure the **OPENMC_CROSS_SECTIONS** environmental variable is set to the
*cross_sections.xml* file in the *data/nndc* directory.
There are two ways to run tests. The first is to use the Makefile present in
the source directory and run the following:
@ -380,11 +381,17 @@ Cross Section Configuration
---------------------------
In order to run a simulation with OpenMC, you will need cross section data for
each nuclide in your problem. Since OpenMC uses ACE format cross sections, 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.
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 multi-group mode, OpenMC utilizes an XML-based library format which can be
used to describe nuclide- or material-specific quantities.
Using ENDF/B-VII.1 Cross Sections from NNDC
-------------------------------------------
@ -399,9 +406,10 @@ extract, and set up a confiuration file:
cd openmc/data
python get_nndc_data.py
At this point, you should set the :envvar:`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.
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.
Using JEFF Cross Sections from OECD/NEA
---------------------------------------
@ -427,8 +435,8 @@ the following steps must be taken:
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:`CROSS_SECTIONS` environment variable to the absolute path of the
``cross_sections.xml`` file.
:envvar:`OPENMC_CROSS_SECTIONS` environment variable to the absolute path of
the ``cross_sections.xml`` file.
Using Cross Sections from MCNP
------------------------------
@ -436,8 +444,9 @@ 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:`CROSS_SECTIONS`
environment variable to the absolute path of the ``cross_sections.xml`` file.
: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.
Using Cross Sections from Serpent
---------------------------------
@ -445,10 +454,21 @@ Using Cross Sections from Serpent
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:`CROSS_SECTIONS`
environment variable to the absolute path of the ``cross_sections_serpent.xml``
: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
--------------------------------
Multi-group cross section libraries are generally tailored to the specific
calculation to be performed. Therefore, at this point in time, OpenMC is not
distributed with any pre-existing multi-group cross section libraries.
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
.. _NNDC: http://www.nndc.bnl.gov/endf/b7.1/acefiles.html
.. _NEA: http://www.oecd-nea.org

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docs/source/usersguide/mgxs_library.rst Normal file
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@ -0,0 +1,302 @@
.. _usersguide_mgxs_library:
========================================
Multi-Group Cross Section Library Format
========================================
OpenMC can be run in continuous-energy mode or multi-group mode, provided the
nuclear data is available. In continuous-energy mode, the
``cross_sections.xml`` file contains necessary meta-data for each data set,
including the name and a file system location where the complete library
can be found. In multi-group mode, this ``cross_sections.xml`` file contains
this same meta-data describing the nuclide or material, but also contains the
group-wise nuclear data. This portion of the manual describes the format of
the multi-group data library required to be used in the ``cross_sections.xml``
file.
Similar to the other input file types, the multi-group library is provided in
the XML_ format. This library must provide some meta-data about the library
itself (such as the number of groups and the group structure, etc.) as well as
the actual cross section data itself for each of the necessary nuclides or
materials.
.. _XML: http://www.w3.org/XML/
------------------------------------------------
MGXS Library Specification -- cross_sections.xml
------------------------------------------------
The multi-group library meta-data is contained within the groups_,
group_structure_, and inverse_velocities_ elements.
The actual multi-group data itself is contained within the xsdata_ element.
.. _groups:
``<groups>`` Element
----------------------------------
The ``<groups>`` element has no attributes and simply provides the number of
energy groups contained within the library.
*Default*: None, this must be provided.
.. _group_structure:
``<group_structure>`` Element
-----------------------------
The ``<group_structure>`` element has no attributes and should be provided as a
monotonically increasing list of bounding energies, in MeV, for a number of
groups. To provide proper energy boundaries, the length of the data within the
``<group_structure>`` element should be one more than the number of groups in
the problem. For example, a two-group problem could be specified as:
.. code-block:: xml
<group_structure> 0.0 0.625E-6 20.0 </group_structure>
*Default*: None, this must be provided.
.. _inverse_velocities:
``<inverse_velocities>`` Element
--------------------------------
The ``<inverse_velocities>`` element optionally indicates the average
inverse velocity corresponding to each of the groups in the problem.
This element should therefore be an array with a length which matches the
number of groups set in the groups_ element.
*Default*: Should this be needed by the presence of an ``inverse-velocity``
score in the ``tallies.xml`` file and not provided in this element, OpenMC
will simply convert the group mid-point energy to an inverse of the velocity
and use this information for tallying.
.. _xsdata:
``<xsdata>`` Element
--------------------
The ``<xsdata>`` element contains the nuclide or material-specific meta-data as
well as the actual cross section data. The following are the
attributes/sub-elements required to describe the meta-data:
:name:
The name of the microscopic or macroscopic data set. An extension to the
name must be provided (e.g., the ``.300K`` in ``UO2.300K``). The name and
extension together must be twelve or less characters in length. This
extension must follow a period and be five characters or less in length.
similar to the equivalent in the continuous-energy ``cross_sections.xml``
file, is used to denote variants of the particular nuclide or material of
interest (i.e. the ``UO2`` data in this example could have been generated
at a temperature of 300K).
*Default*: None, this must be provided.
:alias:
An alternative name to use for the microscopic or macroscopic data set.
*Default*: If no alias is provided, it will adopt the value of ``name``.
:kT:
The temperature times Boltzmann's constant (in units of MeV) at which the
data was generated.
*Default*: Room temperature, 2.53E-8 MeV
:fissionable:
This element states whether or not the data in question is fissionable.
Accepted values are "true" or "false".
*Default*: None, this element must be provided.
:representation:
This element provides the method used to generate and represent the
multi-group cross sections. That is, whether they were generated with
scalar flux weighting (or reduced to an equivalent representation)
and thus are angle-independent, or if the data was generated with angular
dependent fluxes and thus the data is angle-dependent. The options are
either "isotropic" or "angle".
*Default*: "isotropic"
:num_azimuthal:
This element provides the number of equal width angular bins that the
azimuthal angular domain is subdivided in the case of angle-dependent
cross sections (i.e., "angle" is passed to the ``representation`` element).
Note that these bins are equal in azimuthal angle widths, not equal in the
cosine of the azimuthal angle widths.
*Default*: If ``representation`` is "angle", this must be provided. This
parameter is not used for other ``representation`` types.
:num_polar:
This element provides the number of equal width angular bins that the
polar angular domain is subdivided in the case of angle-dependent
cross sections (i.e., "angle" is passed to the ``representation`` element).
Note that these bins are equal in polar angle widths, not equal in the
cosine of the polar angle widths.
*Default*: If ``representation`` is "angle", this must be provided. This
parameter is not used for other ``representation`` types.
:scatt_type:
This element provides the representation of the angular distribution
associated with each group-to-group transfer probability. The options are
either "legendre", "histogram", or "tabular".
The "legendre" option means the angular distribution has been
expanded via Legendre polynomials of the order provided in the "order"
element.
The "histogram" option means the angular distribution is provided in
an equi-width histogram format with a number of bins as provided in the
"order" element. This is useful when the angular distribution was
obtained from a Monte Carlo tally and thus is natively in the histogram
format.
The "tabular" option means the angular distribution is provided in an
equi-spaced point-wise representation.
*Default*: "legendre"
:order:
This element provides either the Legendre order, number of bins, or number
of points used to describe the angular distribution associated with each
group-to-group transfer probability. The specific meaning of this bin
depends upon the value of ``scatt_type`` as discussed above.
*Default*: None, this element must be provided.
:tabular_legendre:
This optional element is used to set how the Legendre scattering kernel, if
provided via the ``scatt_type`` element above, is represented and thus used
during the scattering process. Specifically, the options are to either
convert the Legendre expansion to a tabular representation or leave it as
a set of Legendre coefficients. Converting to a tabular representation will
cost memory but is likely to decrease runtime compared to leaving as a
set of Legendre coefficients. This element has the following
attributes/sub-elements:
:enable:
This attribute/sub-element denotes whether or not the conversion to the
tabular format should be performed or not. A value of "true" means
the conversion should be performed, "false" means it should not.
*Default*: "true"
:num_points:
If the conversion is to take place the number of tabular points is
required. This attribute/sub-element allows the user to set the desired
number of points.
*Default*: 33
The following attributes/sub-elements are the cross section values to
be used during the transport process.
:total:
This element requires the group-wise total cross section ordered by
increasing group index (i.e., fast to thermal). If ``representation`` is
"isotropic", then the length of this list should equal the number of
groups described in the ``groups`` element. If ``representation`` is
"angle", then the length of this list should equal the number of groups
times the number of azimuthal angles times the number of polar angles,
with the inner-dimension being groups, intermediate-dimension being
azimuthal angles and outer-dimension being the polar angles.
*Default*: If not provided, it will be determined by summing the
absorption and scattering cross sections.
:absorption:
This element requires the group-wise absorption cross section ordered by
increasing group index (i.e., fast to thermal). If ``representation`` is
"isotropic", then the length of this list should equal the number of
groups described in the ``groups`` element. If ``representation`` is
"angle", then the length of this list should equal the number of groups
times the number of azimuthal angles times the number of polar angles,
with the inner-dimension being groups, intermediate-dimension being
azimuthal angles and outer-dimension being the polar angles.
*Default*: None, this must be provided.
:scatter:
This element requires the scattering moment matrices presented with the
columns representing incoming group and rows representing the outgoing
group. That is, down-scatter will be above the diagonal of the resultant
matrix. This matrix is repeated for every Legendre order (in order of
increasing orders) if ``scatt_type`` is "legendre"; otherwise, this
matrix is repeated for every bin of the histogram or tabular
representation. Finally, if ``representation`` is "angle", the above
is repeated for every azimuthal angle and every polar angle, in that
order.
*Default*: None, this must be provided.
:multiplicity:
This element provides the ratio of neutrons produced in scattering
collisions to the neutrons which undergo scattering collisions; that is,
the multiplicity provides the code with a scaling factor to account for
neutrons being produced in (n,xn) reactions. This information is assumed
isotropic and therefore does not need to be repeated for every Legendre
moment or histogram/tabular bin. This matrix follows the same arrangement
as described for the ``scatter`` element, with the exception of the
data needed to provide the scattering type information.
*Default*: Multiplicities of 1.0 are assumed (i.e., (n,xn) reactions are
neglected).
The following fission-specific data are only needed should ``fissionable``
be "true".
:fission:
This element requires the group-wise fission cross section ordered by
increasing group index (i.e., fast to thermal). If ``representation`` is
"isotropic", then the length of this list should equal the number of
groups described in the ``groups`` element. If ``representation`` is
"angle", then the length of this list should equal the number of groups
times the number of azimuthal angles times the number of polar angles,
with the inner-dimension being groups, intermediate-dimension being
azimuthal angles and outer-dimension being the polar angles.
*Default*: None, this is required only if fission tallies are
requested and the material is fissionable.
:kappa_fission:
This element requires the group-wise kappa-fission cross section ordered by
increasing group index (i.e., fast to thermal). If ``representation`` is
"isotropic", then the length of this list should equal the number of
groups described in the ``groups`` element. If ``representation`` is
"angle", then the length of this list should equal the number of groups
times the number of azimuthal angles times the number of polar angles,
with the inner-dimension being groups, intermediate-dimension being
azimuthal angles and outer-dimension being the polar angles.
*Default*: None, this is required only if kappa_fission tallies are
requested and the material is fissionable.
:chi:
This element requires the group-wise fission spectra ordered by
increasing group index (i.e., fast to thermal). This element should be
used if making the common approximation that the fission spectra does
not depend on incoming energy. If the user does not wish to make this
approximation, then this should not be provided and this information
included in the ``nu_fission`` element instead. If ``representation`` is
"isotropic", then the length of this list should equal the number of
groups described in the ``groups`` element. If ``representation`` is
"angle", then the length of this list should equal the number of groups
times the number of azimuthal angles times the number of polar angles,
with the inner-dimension being groups, intermediate-dimension being
azimuthal angles and outer-dimension being the polar angles.
*Default*: None, either this element is provided or ``nu_fission`` is
provided in fission matrix form, or the material is not fissionable.
:nu_fission:
This element provides either the group-wise fission production cross
section vector (i.e., if ``chi`` is provided), or is the group-wise fission
production matrix. If providing the vector, it should be ordered the same
as the ``fission`` data. If providing the matrix, it should be ordered
the same as the ``multiplicity`` matrix.
*Default*: None, either this element must be provided if the material
is fissionable.

3
docs/source/usersguide/output/particle_restart.rst
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@ -46,7 +46,8 @@ The current revision of the particle restart file format is 1.
**/energy** (*double*)
Energy of the particle in MeV.
Energy of the particle in MeV for continuous-energy mode, or the energy
group of the particle for multi-group mode.
**/xyz** (*double[3]*)

5
docs/source/usersguide/output/source.rst
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@ -15,5 +15,6 @@ is that documented here.
**/source_bank** (Compound type)
Source bank information for each particle. The compound type has fields
``wgt``, ``xyz``, ``uvw``, and ``E`` which represent the weight, position,
direction, and energy of the source particle, respectively.
``wgt``, ``xyz``, ``uvw``, ``E``, and ``delayed_group``, which
represent the weight, position, direction, energy, energy group, and
delayed_group of the source particle, respectively.

81
docs/source/usersguide/output/statepoint.rst
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@ -4,7 +4,7 @@
State Point File Format
=======================
The current revision of the statepoint file format is 14.
The current revision of the statepoint file format is 15.
**/filetype** (*char[]*)
@ -39,6 +39,12 @@ The current revision of the statepoint file format is 14.
Pseudo-random number generator seed.
**/run_CE** (*int*)
Flag to denote continuous-energy or multi-group mode. A value of 1
indicates a continuous-energy run while a value of 0 indicates a
multi-group run.
**/run_mode** (*char[]*)
Run mode used. A value of 1 indicates a fixed-source run and a value of 2
@ -258,7 +264,7 @@ if run_mode == 'k-eigenvalue':
Accumulated sum and sum-of-squares for each global tally. The compound type
has fields named ``sum`` and ``sum_sq``.
**tallies_present** (*int*)
**/tallies_present** (*int*)
Flag indicated if tallies are present in the file.
@ -267,5 +273,72 @@ if (run_mode == 'k-eigenvalue' and source_present > 0)
**/source_bank** (Compound type)
Source bank information for each particle. The compound type has fields
``wgt``, ``xyz``, ``uvw``, and ``E`` which represent the weight,
position, direction, and energy of the source particle, respectively.
``wgt``, ``xyz``, ``uvw``, ``E``, ``g``, and ``delayed_group``, which
represent the weight, position, direction, energy, energy group, and
delayed_group of the source particle, respectively.
**/runtime/total initialization** (*double*)
Time (in seconds on the master process) spent reading inputs, allocating
arrays, etc.
**/runtime/reading cross sections** (*double*)
Time (in seconds on the master process) spent loading cross section
libraries (this is a subset of initialization).
**/runtime/simulation** (*double*)
Time (in seconds on the master process) spent between initialization and
finalization.
**/runtime/transport** (*double*)
Time (in seconds on the master process) spent transporting particles.
**/runtime/inactive batches** (*double*)
Time (in seconds on the master process) spent in the inactive batches
(including non-transport activities like communcating sites).
**/runtime/active batches** (*double*)
Time (in seconds on the master process) spent in the active batches
(including non-transport activities like communicating sites).
**/runtime/synchronizing fission bank** (*double*)
Time (in seconds on the master process) spent sampling source particles
from fission sites and communicating them to other processes for load
balancing.
**/runtime/sampling source sites** (*double*)
Time (in seconds on the master process) spent sampling source particles
from fission sites.
**/runtime/SEND-RECV source sites** (*double*)
Time (in seconds on the master process) spent communicating source sites
between processes for load balancing.
**/runtime/accumulating tallies** (*double*)
Time (in seconds on the master process) spent communicating tally results
and evaluating their statistics.
**/runtime/CMFD** (*double*)
Time (in seconds on the master process) spent evaluating CMFD.
**/runtime/CMFD building matrices** (*double*)
Time (in seconds on the master process) spent buliding CMFD matrices.
**/runtime/CMFD solving matrices** (*double*)
Time (in seconds on the master process) spent solving CMFD matrices.
**/runtime/total** (*double*)
Total time spent (in seconds on the master process) in the program.

7
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@ -297,6 +297,13 @@ The current revision of the summary file format is 1.
Filter offset (used for distribcell filter).
**/tallies/tally <uid>/filter <j>/paths** (*char[][]*)
The paths traversed through the CSG tree to reach each distribcell
instance (for 'distribcell' filters only). This consists of the integer
IDs for each universe, cell and lattice delimited by '->'. Each lattice
cell is specified by its (x,y) or (x,y,z) indices.
**/tallies/tally <uid>/filter <j>/n_bins** (*int*)
Number of bins for the j-th filter.

8
docs/source/usersguide/processing.rst
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@ -196,10 +196,10 @@ Data Extraction
A great deal of information is available in statepoint files (See
:ref:`usersguide_statepoint`), all of which is accessible through the Python
API. The ``openmc.statepoint`` module (see :ref:`pythonapi_statepoint`) provides
a class to load statepoints and access data as requested; it is used in many of
the provided plotting utilities, OpenMC's regression test suite, and can be used
in user-created scripts to carry out manipulations of the data.
API. The :class:`openmc.StatePoint` class can load statepoints and access data
as requested; it is used in many of the provided plotting utilities, OpenMC's
regression test suite, and can be used in user-created scripts to carry out
manipulations of the data.
An :ref:`example IPython notebook <notebook_post_processing>` demonstrates how
to extract data from a statepoint using the Python API.