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61
docs/source/capi/index.rst
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@ -46,10 +46,13 @@ Type Definitions
Functions
---------
.. c:function:: void openmc_calculate_volumes()
.. c:function:: int openmc_calculate_volumes()
Run a stochastic volume calculation
:return: Return status (negative if an error occurred)
:rtype: int
.. c:function:: int openmc_cell_get_fill(int32_t index, int* type, int32_t** indices, int32_t* n)
Get the fill for a cell
@ -192,11 +195,14 @@ Functions
:return: Return status (negative if an error occurred)
:rtype: int
.. c:function:: void openmc_finalize()
.. c:function:: int openmc_finalize()
Finalize a simulation
.. c:function:: void openmc_find(double* xyz, int rtype, int32_t* id, int32_t* instance)
:return: Return status (negative if an error occurs)
:rtype: int
.. c:function:: int openmc_find(double* xyz, int rtype, int32_t* id, int32_t* instance)
Determine the ID of the cell/material containing a given point
@ -207,6 +213,8 @@ Functions
occurs, the ID is -1.
:param int32_t* instance: If a cell is repeated in the geometry, the instance
of the cell that was found and zero otherwise.
:return: Return status (negative if an error occurs)
:rtype: int
.. c:function:: int openmc_get_cell_index(int32_t id, int32_t* index)
@ -247,11 +255,12 @@ Functions
:return: Return status (negative if an error occurs)
:rtype: int
.. c:function:: int openmc_get_nuclide_index(char name[], int* index)
.. c:function:: int openmc_get_nuclide_index(const char name[], int* index)
Get the index in the nuclides array for a nuclide with a given name
:param char[] name: Name of the nuclide
:param name: Name of the nuclide
:type name: const char[]
:param int* index: Index in the nuclides array
:return: Return status (negative if an error occurs)
:rtype: int
@ -265,17 +274,24 @@ Functions
:return: Return status (negative if an error occurs)
:rtype: int
.. c:function:: void openmc_hard_reset()
.. c:function:: int openmc_hard_reset()
Reset tallies, timers, and pseudo-random number generator state
.. c:function:: void openmc_init(const int* intracomm)
:return: Return status (negative if an error occurs)
:rtype: int
.. c:function:: int openmc_init(int argc, char** argv, const void* intracomm)
Initialize OpenMC
:param int argc: Number of command-line arguments (including command)
:param char** argv: Command-line arguments
:param intracomm: MPI intracommunicator. If MPI is not being used, a null
pointer should be passed.
:type intracomm: const int*
:type intracomm: const void*
:return: Return status (negative if an error occurs)
:rtype: int
.. c:function:: int openmc_load_nuclide(char name[])
@ -393,26 +409,41 @@ Functions
:return: Return status (negative if an error occurs)
:rtype: int
.. c:function:: void openmc_plot_geometry()
.. c:function:: int openmc_plot_geometry()
Run plotting mode.
.. c:function:: void openmc_reset()
:return: Return status (negative if an error occurs)
:rtype: int
.. c:function:: int openmc_reset()
Resets all tally scores
.. c:function:: void openmc_run()
:return: Return status (negative if an error occurs)
:rtype: int
.. c:function:: int openmc_run()
Run a simulation
.. c:function:: void openmc_simulation_finalize()
:return: Return status (negative if an error occurs)
:rtype: int
.. c:function:: int openmc_simulation_finalize()
Finalize a simulation.
.. c:function:: void openmc_simulation_init()
:return: Return status (negative if an error occurs)
:rtype: int
.. c:function:: int openmc_simulation_init()
Initialize a simulation. Must be called after openmc_init().
:return: Return status (negative if an error occurs)
:rtype: int
.. c:function:: int openmc_source_bank(struct Bank** ptr, int64_t* n)
Return a pointer to the source bank array.
@ -432,13 +463,15 @@ Functions
:return: Return status (negative if an error occurred)
:rtype: int
.. c:function:: void openmc_statepoint_write(const char filename[])
.. c:function:: int openmc_statepoint_write(const char filename[])
Write a statepoint file
:param filename: Name of file to create. If a null pointer is passed, a
filename is assigned automatically.
:type filename: const char[]
:return: Return status (negative if an error occurs)
:rtype: int
.. c:function:: int openmc_tally_get_id(int32_t index, int32_t* id)

7
docs/source/devguide/styleguide.rst
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@ -219,8 +219,8 @@ Curly braces
For a function definition, the opening and closing braces should each be on
their own lines. This helps distinguish function code from the argument list.
If the entire function fits on one line, then the braces can be on the same
line. e.g.:
If the entire function fits on one or two lines, then the braces can be on the
same line. e.g.:
.. code-block:: C++
@ -238,6 +238,9 @@ line. e.g.:
int return_one() {return 1;}
int return_one()
{return 1;}
For a conditional, the opening brace should be on the same line as the end of
the conditional statement. If there is a following ``else if`` or ``else``
statement, the closing brace should be on the same line as that following

13
docs/source/examples/expansion-filters.rst Normal file
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@ -0,0 +1,13 @@
.. _notebook_expansion:
=====================
Functional Expansions
=====================
.. only:: html
.. notebook:: ../../../examples/jupyter/expansion-filters.ipynb
.. only:: latex
IPython notebooks must be viewed in the online HTML documentation.

14
docs/source/examples/index.rst
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@ -1,13 +1,12 @@
.. _examples:
=================
Example Notebooks
=================
========
Examples
========
The following series of Jupyter_ Notebooks provide examples for usage of OpenMC
features via the :ref:`pythonapi`.
.. _Jupyter: https://jupyter.org/
The following series of `Jupyter <https://jupyter.org/>`_ Notebooks provide
examples for how to use various features of OpenMC by leveraging the
:ref:`pythonapi`.
-----------
Basic Usage
@ -20,6 +19,7 @@ Basic Usage
post-processing
pandas-dataframes
tally-arithmetic
expansion-filters
search
triso
candu

25
docs/source/io_formats/data_wmp.rst
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@ -28,12 +28,6 @@ Windowed Multipole Library Format
":math:`r`" and ":math:`i`" identifiers, similar to how `h5py`_ does it.
- **end_E** (*double*)
Highest energy the windowed multipole part of the library is valid for.
- **energy_points** (*double[]*)
Energy grid for the pointwise library in the reaction group.
- **fissionable** (*int*)
1 if this nuclide has fission data. 0 if it does not.
- **fit_order** (*int*)
The order of the curve fit.
- **formalism** (*int*)
The formalism of the underlying data. Uses the `ENDF-6`_ format
formalism numbers.
@ -51,18 +45,6 @@ Windowed Multipole Library Format
- **l_value** (*int[]*)
The index for a corresponding pole. Equivalent to the :math:`l` quantum
number of the resonance the pole comes from :math:`+1`.
- **length** (*int*)
Total count of poles in `data`.
- **max_w** (*int*)
Maximum number of poles in a window.
- **MT_count** (*int*)
Number of pointwise tables in the library.
- **MT_list** (*int[]*)
A list of available MT identifiers. See `ENDF-6`_ for meaning.
- **n_grid** (*int*)
Total length of the pointwise data.
- **num_l** (*int*)
Number of possible :math:`l` quantum states for this nuclide.
- **pseudo_K0RS** (*double[]*)
:math:`l` dependent value of
@ -90,13 +72,6 @@ Windowed Multipole Library Format
The pole to start from for each window.
- **w_end** (*int[]*)
The pole to end at for each window.
- **windows** (*int*)
Number of windows.
**/nuclide/reactions/MT<i>**
- **MT_sigma** (*double[]*) -- Cross section value for this reaction.
- **Q_value** (*double*) -- Energy released in this reaction, in eV.
- **threshold** (*int*) -- The first non-zero entry in ``MT_sigma``.
.. _h5py: http://docs.h5py.org/en/latest/
.. _ENDF-6: https://www.oecd-nea.org/dbdata/data/manual-endf/endf102.pdf

87
docs/source/io_formats/settings.rst
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@ -93,29 +93,13 @@ or ``multi-group``.
*Default*: continuous-energy
---------------------
``<entropy>`` Element
---------------------
--------------------------
``<entropy_mesh>`` Element
--------------------------
The ``<entropy>`` element describes a mesh that is used for calculating Shannon
entropy. This mesh should cover all possible fissionable materials in the
problem. It has the following attributes/sub-elements:
:dimension:
The number of mesh cells in the x, y, and z directions, respectively.
*Default*: If this tag is not present, the number of mesh cells is
automatically determined by the code.
:lower_left:
The Cartesian coordinates of the lower-left corner of the mesh.
*Default*: None
:upper_right:
The Cartesian coordinates of the upper-right corner of the mesh.
*Default*: None
The ``<entropy_mesh>`` element indicates the ID of a mesh that is to be used for
calculating Shannon entropy. The mesh should cover all possible fissionable
materials in the problem and is specified using a :ref:`mesh_element`.
-----------------------------------
``<generations_per_batch>`` Element
@ -199,6 +183,36 @@ then, OpenMC will only use up to the :math:`P_1` data.
.. note:: This element is not used in the continuous-energy
:ref:`energy_mode`.
.. _mesh_element:
------------------
``<mesh>`` Element
------------------
The ``<mesh>`` element describes a mesh that is used either for calculating
Shannon entropy, applying the uniform fission site method, or in tallies. For
Shannon entropy meshes, the mesh should cover all possible fissionable materials
in the problem. It has the following attributes/sub-elements:
:id:
A unique integer that is used to identify the mesh.
:dimension:
The number of mesh cells in the x, y, and z directions, respectively.
*Default*: If this tag is not present, the number of mesh cells is
automatically determined by the code.
:lower_left:
The Cartesian coordinates of the lower-left corner of the mesh.
*Default*: None
:upper_right:
The Cartesian coordinates of the upper-right corner of the mesh.
*Default*: None
-----------------------
``<no_reduce>`` Element
-----------------------
@ -765,30 +779,15 @@ has the following attributes/sub-elements:
------------------------
``<uniform_fs>`` Element
``<ufs_mesh>`` Element
------------------------
The ``<uniform_fs>`` element describes a mesh that is used for re-weighting
source sites at every generation based on the uniform fission site methodology
described in Kelly et al., "MC21 Analysis of the Nuclear Energy Agency Monte
Carlo Performance Benchmark Problem," Proceedings of *Physor 2012*, Knoxville,
TN (2012). This mesh should cover all possible fissionable materials in the
problem. It has the following attributes/sub-elements:
:dimension:
The number of mesh cells in the x, y, and z directions, respectively.
*Default*: None
:lower_left:
The Cartesian coordinates of the lower-left corner of the mesh.
*Default*: None
:upper_right:
The Cartesian coordinates of the upper-right corner of the mesh.
*Default*: None
The ``<ufs_mesh>`` element indicates the ID of a mesh that is used for
re-weighting source sites at every generation based on the uniform fission site
methodology described in Kelly et al., "MC21 Analysis of the Nuclear Energy
Agency Monte Carlo Performance Benchmark Problem," Proceedings of *Physor 2012*,
Knoxville, TN (2012). The mesh should cover all possible fissionable materials
in the problem and is specified using a :ref:`mesh_element`.
.. _verbosity:

16
docs/source/io_formats/source.rst
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@ -8,13 +8,13 @@ Normally, source data is stored in a state point file. However, it is possible
to request that the source be written separately, in which case the format used
is that documented here.
**/filetype** (*char[]*)
**/**
String indicating the type of file.
:Attributes: - **filetype** (*char[]*) -- String indicating the type of file.
**/source_bank** (Compound type)
Source bank information for each particle. The compound type has fields
``wgt``, ``xyz``, ``uvw``, ``E``, and ``delayed_group``, which
represent the weight, position, direction, energy, energy group, and
delayed_group of the source particle, respectively.
:Datasets:
- **source_bank** (Compound type) -- Source bank information for each
particle. The compound type has fields ``wgt``, ``xyz``, ``uvw``,
``E``, and ``delayed_group``, which represent the weight, position,
direction, energy, energy group, and delayed_group of the source
particle, respectively.

9
docs/source/io_formats/statepoint.rst
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@ -133,15 +133,8 @@ The current version of the statepoint file format is 17.0.
- **derivative** (*int*) -- ID of the derivative applied to the
tally.
- **n_score_bins** (*int*) -- Number of scoring bins for a single
nuclide. In general, this can be greater than the number of
user-specified scores since each score might have multiple scoring
bins, e.g., scatter-PN.
nuclide.
- **score_bins** (*char[][]*) -- Values of specified scores.
- **n_user_scores** (*int*) -- Number of scores without accounting
for those added by expansions, e.g. scatter-PN.
- **moment_orders** (*char[][]*) -- Tallying moment orders for
Legendre and spherical harmonic tally expansions (e.g., 'P2',
'Y1,2', etc.).
- **results** (*double[][][2]*) -- Accumulated sum and sum-of-squares
for each bin of the i-th tally. The first dimension represents
combinations of filter bins, the second dimensions represents

16
docs/source/io_formats/summary.rst
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@ -4,7 +4,7 @@
Summary File Format
===================
The current version of the summary file format is 5.0.
The current version of the summary file format is 6.0.
**/**
@ -104,8 +104,13 @@ The current version of the summary file format is 5.0.
- **atom_density** (*double[]*) -- Total atom density of the material
in atom/b-cm.
- **nuclides** (*char[][]*) -- Array of nuclides present in the
material, e.g., 'U235'.
material, e.g., 'U235'. This data set is only present if nuclides
are used.
- **nuclide_densities** (*double[]*) -- Atom density of each nuclide.
This data set is only present if 'nuclides' data set is present.
- **macroscopics** (*char[][]*) -- Array of macroscopic data sets
present in the material. This dataset is only present if
macroscopic data sets are used in multi-group mode.
- **sab_names** (*char[][]*) -- Names of
S(:math:`\alpha,\beta`) tables assigned to the material.
@ -116,6 +121,13 @@ The current version of the summary file format is 5.0.
:Datasets: - **names** (*char[][]*) -- Names of nuclides.
- **awrs** (*float[]*) -- Atomic weight ratio of each nuclide.
**/macroscopics/**
:Attributes: - **n_macroscopics** (*int*) -- Number of macroscopic data sets
in the problem.
:Datasets: - **names** (*char[][]*) -- Names of the macroscopic data sets.
**/tallies/tally <uid>/**
:Datasets: - **name** (*char[]*) -- Name of the tally.

15
docs/source/publications.rst
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@ -57,6 +57,11 @@ Benchmarking
Coupling and Multi-physics
--------------------------
- Jun Chen, Liangzhi Cao, Chuanqi Zhao, and Zhouyu Liu, "`Development of
Subchannel Code SUBSC for high-fidelity multi-physics coupling application
<https://doi.org/10.1016/j.egypro.2017.08.121>`_", Energy Procedia, **127**,
264-274 (2017).
- Tianliang Hu, Liangzhu Cao, Hongchun Wu, Xianan Du, and Mingtao He, "`Coupled
neutrons and thermal-hydraulics simulation of molten salt reactors based on
OpenMC/TANSY <https://doi.org/10.1016/j.anucene.2017.05.002>`_,"
@ -98,6 +103,11 @@ Coupling and Multi-physics
Geometry and Visualization
--------------------------
- Jin-Yang Li, Long Gu, Hu-Shan Xu, Nadezha Korepanova, Rui Yu, Yan-Lei Zhu, and
Chang-Ping Qin, "`CAD modeling study on FLUKA and OpenMC for accelerator
driven system simulation <https://doi.org/10.1016/j.anucene.2017.12.050>`_",
*Ann. Nucl. Energy*, **114**, 329-341 (2018).
- Logan Abel, William Boyd, Benoit Forget, and Kord Smith, "Interactive
Visualization of Multi-Group Cross Sections on High-Fidelity Spatial Meshes,"
*Trans. Am. Nucl. Soc.*, **114**, 391-394 (2016).
@ -114,6 +124,11 @@ Geometry and Visualization
Miscellaneous
-------------
- Bruno Merk, Dzianis Litskevich, R. Gregg, and A. R. Mount, "`Demand driven
salt clean-up in a molten salt fast reactor -- Defining a priority list
<https://doi.org/10.1371/journal.pone.0192020>`_", *PLOS One*, **13**,
e0192020 (2018).
- Adam G. Nelson, Samuel Shaner, William Boyd, and Paul K. Romano,
"Incorporation of a Multigroup Transport Capability in the OpenMC Monte Carlo
Particle Transport Code," *Trans. Am. Nucl. Soc.*, **117**, 679-681 (2017).

5
docs/source/pythonapi/base.rst
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@ -109,6 +109,7 @@ Constructing Tallies
openmc.CellbornFilter
openmc.SurfaceFilter
openmc.MeshFilter
openmc.MeshSurfaceFilter
openmc.EnergyFilter
openmc.EnergyoutFilter
openmc.MuFilter
@ -117,6 +118,10 @@ Constructing Tallies
openmc.DistribcellFilter
openmc.DelayedGroupFilter
openmc.EnergyFunctionFilter
openmc.LegendreFilter
openmc.SpatialLegendreFilter
openmc.SphericalHarmonicsFilter
openmc.ZernikeFilter
openmc.Mesh
openmc.Trigger
openmc.TallyDerivative

3
docs/source/pythonapi/capi.rst
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@ -44,5 +44,8 @@ Classes
EnergyFilter
MaterialFilter
Material
Mesh
MeshFilter
MeshSurfaceFilter
Nuclide
Tally

12
docs/source/usersguide/install.rst
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@ -141,16 +141,10 @@ Prerequisites
recommend that your HDF5 installation be built with parallel I/O
features. An example of configuring HDF5_ is listed below::
FC=mpifort ./configure --enable-fortran --enable-parallel
FC=mpifort ./configure --enable-parallel
You may omit ``--enable-parallel`` if you want to compile HDF5_ in serial.
.. important::
If you are building HDF5 version 1.8.x or earlier, you must include
``--enable-fortran2003`` when configuring HDF5 or else OpenMC will not
be able to compile.
.. admonition:: Optional
:class: note
@ -416,7 +410,9 @@ Prerequisites
The Python API works with Python 3.4+. In addition to Python itself, the API
relies on a number of third-party packages. All prerequisites can be installed
using Conda_ (recommended), pip_, or through the package manager in most Linux
distributions.
distributions. To run simulations in parallel using MPI, it is recommended to
build mpi4py, HDF5, h5py from source, in that order, using the same compilers
as for OpenMC.
.. admonition:: Required
:class: error

14
docs/source/usersguide/materials.rst
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@ -43,14 +43,22 @@ of an element, you specify the element itself. For example,
Internally, OpenMC stores data on the atomic masses and natural abundances of
all known isotopes and then uses this data to determine what isotopes should be
added to the material. When the material is later exported to XML for use by the
:ref:`scripts_openmc` executable, you'll see that any natural elements are
:ref:`scripts_openmc` executable, you'll see that any natural elements were
expanded to the naturally-occurring isotopes.
The :meth:`Material.add_element` method can also be used to add uranium at a
specified enrichment through the `enrichment` argument. For example, the
following would add 3.2% enriched uranium to a material::
mat.add_element('U', 1.0, enrichment=3.2)
In addition to U235 and U238, concentrations of U234 and U236 will be present
and are determined through a correlation based on measured data.
Often, cross section libraries don't actually have all naturally-occurring
isotopes for a given element. For example, in ENDF/B-VII.1, cross section
evaluations are given for O16 and O17 but not for O18. If OpenMC is aware of
what cross sections you will be using (either through the
:attr:`Materials.cross_sections` attribute or the
what cross sections you will be using (through the
:envvar:`OPENMC_CROSS_SECTIONS` environment variable), it will attempt to only
put isotopes in your model for which you have cross section data. In the case of
oxygen in ENDF/B-VII.1, the abundance of O18 would end up being lumped with O16.

67
docs/source/usersguide/tallies.rst
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@ -21,7 +21,12 @@ region of phase space, as in:
Thus, to specify a tally, we need to specify what regions of phase space should
be included when deciding whether to score an event as well as what the scoring
function (:math:`f` in the above equation) should be used. The regions of phase
space are called *filters* and the scoring functions are simply called *scores*.
space are generally called *filters* and the scoring functions are simply
called *scores*.
The only cases when filters do not correspond directly with the regions of
phase space are when expansion functions are applied in the integrand, such as
for Legendre expansions of the scattering kernel.
-------
Filters
@ -69,10 +74,9 @@ Scores
------
To specify the scoring functions, a list of strings needs to be given to the
:attr:`Tally.scores` attribute. You can score the flux ('flux'), a reaction rate
('total', 'fission', etc.), or even scattering moments (e.g., 'scatter-P3'). For
example, to tally the elastic scattering rate and the fission neutron
production, you'd assign::
:attr:`Tally.scores` attribute. You can score the flux ('flux'), or a reaction
rate ('total', 'fission', etc.). For example, to tally the elastic scattering
rate and the fission neutron production, you'd assign::
tally.scores = ['elastic', 'nu-fission']
@ -98,12 +102,6 @@ The following tables show all valid scores:
+======================+===================================================+
|flux |Total flux. |
+----------------------+---------------------------------------------------+
|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. |
+----------------------+---------------------------------------------------+
.. table:: **Reaction scores: units are reactions per source particle.**
@ -118,43 +116,10 @@ The following tables show all valid scores:
+----------------------+---------------------------------------------------+
|fission |Total fission reaction rate. |
+----------------------+---------------------------------------------------+
|scatter |Total scattering rate. Can also be identified with |
| |the "scatter-0" response type. |
+----------------------+---------------------------------------------------+
|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>``. |
+----------------------+---------------------------------------------------+
|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>``. |
+----------------------+---------------------------------------------------+
|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 |Total scattering rate. |
+----------------------+---------------------------------------------------+
|total |Total reaction rate. |
+----------------------+---------------------------------------------------+
|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. |
+----------------------+---------------------------------------------------+
|(n,2nd) |(n,2nd) reaction rate. |
+----------------------+---------------------------------------------------+
|(n,2n) |(n,2n) reaction rate. |
@ -248,10 +213,10 @@ The following tables show all valid scores:
+----------------------+---------------------------------------------------+
|nu-fission |Total production of neutrons due to fission. |
+----------------------+---------------------------------------------------+
|nu-scatter, |These scores are similar in functionality to their |
|nu-scatter-N, |``scatter*`` equivalents except the total |
|nu-scatter-PN, |production of neutrons due to scattering is scored |
|nu-scatter-YN |vice simply the scattering rate. This accounts for |
|nu-scatter, |This score is similar in functionality to the |
| |``scatter`` score 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. |
+----------------------+---------------------------------------------------+
@ -261,7 +226,7 @@ The following tables show all valid scores:
+----------------------+---------------------------------------------------+
|Score | Description |
+======================+===================================================+
|current |Used in combination with a mesh filter: |
|current |Used in combination with a meshsurface filter: |
| |Partial currents on the boundaries of each cell in |
| |a mesh. It may not be used in conjunction with any |
| |other score. Only energy and mesh filters may be |
@ -269,7 +234,7 @@ The following tables show all valid scores:
| |Used in combination with a surface filter: |
| |Net currents on any surface previously defined in |
| |the geometry. It may be used along with any other |
| |filter, except mesh filters. |
| |filter, except meshsurface filters. |
| |Surfaces can alternatively be defined with cell |
| |from and cell filters thereby resulting in tallying|
| |partial currents. |