From aa1f4c66ab2e68258c2c3dbc84af57c455f2d47e Mon Sep 17 00:00:00 2001 From: Adam Nelson Date: Mon, 7 Apr 2014 14:27:11 -0400 Subject: [PATCH] Updated documentation for the new score types.; --- docs/source/usersguide/input.rst | 162 ++++++++++++++++++------------- 1 file changed, 93 insertions(+), 69 deletions(-) diff --git a/docs/source/usersguide/input.rst b/docs/source/usersguide/input.rst index eebe6bfc1..ef4b658de 100644 --- a/docs/source/usersguide/input.rst +++ b/docs/source/usersguide/input.rst @@ -41,7 +41,7 @@ for the geometry, materials, and settings. Additionally, there are three optiona input files. The first is a tallies XML file that specifies physical quantities to be tallied. The second is a plots XML file that specifies regions of geometry which should be plotted. The third is a CMFD XML file that specifies coarse mesh -acceleration geometry and execution parameters. OpenMC expects that these +acceleration geometry and execution parameters. OpenMC expects that these files are called: * ``geometry.xml`` @@ -105,7 +105,7 @@ default. This element has the following attributes/sub-elements: The ```` element indicates that a :math:`k`-eigenvalue calculation should be performed. It has the following attributes/sub-elements: - :batches: + :batches: The total number of batches, where each batch corresponds to multiple fission source iterations. Batching is done to eliminate correlation between realizations of random variables. @@ -171,7 +171,7 @@ problem. It has the following attributes/sub-elements: The ```` element indicates that a fixed source calculation should be performed. It has the following attributes/sub-elements: - :batches: + :batches: The total number of batches. For fixed source calculations, each batch represents a realization of random variables for tallies. @@ -206,7 +206,7 @@ out the file and "false" will not. *Default*: false - :summary: + :summary: Writes out an ASCII summary file describing all of the user input files that were read in. @@ -274,7 +274,7 @@ attributes/sub-elements: An element specifying the spatial distribution of source sites. This element has the following attributes: - :type: + :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 @@ -298,7 +298,7 @@ attributes/sub-elements: An element specifying the angular distribution of source sites. This element has the following attributes: - :type: + :type: The type of angular distribution. Valid options are "isotropic" and "monodirectional". The angle of the particle emitted from a source site is isotropic if the "isotropic" option is given. The angle of the particle @@ -321,7 +321,7 @@ attributes/sub-elements: An element specifying the energy distribution of source sites. This element has the following attributes: - :type: + :type: The type of energy distribution. Valid options are "monoenergetic", "watt", and "maxwell". The "monoenergetic" option produces source sites at @@ -350,8 +350,8 @@ attributes/sub-elements: The ```` element indicates at what batches a state point file should be written. A state point file can be used to restart a run or to get -tally results at any batch. The default behavior when using this tag is to -write out the source bank in the state_point file. This behavior can be +tally results at any batch. The default behavior when using this tag is to +write out the source bank in the state_point file. This behavior can be customized by using the ```` element. This element has the following attributes/sub-elements: @@ -371,8 +371,8 @@ following attributes/sub-elements: ```` Element -------------------------- -The ```` element indicates at what batches the source bank -should be written. The source bank can be either written out within a state +The ```` element indicates at what batches the source bank +should be written. The source bank can be either written out within a state point file or separately in a source point file. This element has the following attributes/sub-elements: @@ -401,17 +401,17 @@ attributes/sub-elements: :source_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 + 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. *Default*: true :overwrite_latest: If this element is set to "true", a source point file containing - the source bank will be written out to a separate file named - ``source.binary`` or ``source.h5`` depending on if HDF5 is enabled. + the source bank will be written out to a separate file named + ``source.binary`` or ``source.h5`` depending on if HDF5 is enabled. This file will be overwritten at every single batch so that the latest - source bank will be available. It should be noted that a user can set both + source bank will be available. It should be noted that a user can set both this element to "true" and specify batches to write a permanent source bank. *Default*: false @@ -845,7 +845,7 @@ The ```` 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 + for output purposes. This string is limited to 52 characters for formatting purposes. :filter: @@ -915,10 +915,11 @@ The ```` element accepts the following sub-elements: :scores: A space-separated list of the desired responses to be accumulated. Accepted - options are "flux", "total", "scatter", "nu-scatter", "scatter-N", - "scatter-PN", "absorption", "fission", "nu-fission", "kappa-fission", - "current", and "events". These corresponding to the following physical - quantities. + options are "flux", "total", "scatter", "absorption", "fission", + "nu-fission", "kappa-fission", "nu-scatter", "scatter-N", "scatter-PN", + "scatter-YN", "nu-scatter-N", "nu-scatter-PN", "nu-scatter-YN", "flux-YN", + "total-YN", "current", and "events". These corresponding to the following + physical quantities: :flux: Total flux @@ -930,24 +931,6 @@ The ```` element accepts the following sub-elements: Total scattering rate. Can also be identified with the ``scatter-0`` response type. - :nu-scatter: - Total production of neutrons due to scattering. This accounts for - multiplicity from (n,2n), (n,3n), and (n,4n) reactions and should be - slightly higher than the scattering rate. - - :scatter-N: - Tally the N\ :sup:`th` \ scattering moment, where N is the Legendre - expansion order. N must be between 0 and 10. As an example, tallying the - 2\ :sup:`nd` \ scattering moment would be specified as `` - scatter-2 ``. - - :scatter-PN: - Tally all of the scattering moments from order 0 to N, where N is the - Legendre expansion order. That is, ``scatter-P1`` is equivalent to - requesting tallies of ``scatter-0`` and ``scatter-1``. N must be between - 0 and 10. As an example, tallying up to the 2\ :sup:`nd` \ scattering - moment would be specified as `` scatter-P2 ``. - :absorption: Total absorption rate. This accounts for all reactions which do not produce secondary neutrons. @@ -957,7 +940,7 @@ The ```` element accepts the following sub-elements: :nu-fission: Total production of neutrons due to fission - + :kappa-fission: The recoverable energy production rate due to fission. The recoverable energy is defined as the fission product kinetic energy, prompt and @@ -967,6 +950,47 @@ The ```` element accepts the following sub-elements: prompt and delayed :math:`\gamma`-rays are assumed to deposit their energy locally. + :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 + `` scatter-2 ``. + + :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 + `` scatter-P2 ``. + + :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. + + :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. + + :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. + + :total-YN: + Spherical harmonic expansion of the incoming particle's 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. + :current: Partial currents on the boundaries of each cell in a mesh. @@ -1137,7 +1161,7 @@ attributes or sub-elements. These are not used in "voxel" plots: Any number of this optional tag may be included in each ```` element, which can override the default random colors for cells or materials. Each ``col_spec`` element must contain ``id`` and ``rgb`` sub-elements. - + :id: Specifies the cell or material unique id for the color specification. @@ -1174,20 +1198,20 @@ 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 ```` element in +Currently, it allows users to accelerate fission source convergence during +inactive neutron batches. To run CMFD, the ```` element in ``settings.xml`` should be set to "true". ```` Element -------------------------- The ```` element controls the batch where CMFD tallies should be -reset. CMFD tallies should be reset before active batches so they are accumulated +reset. CMFD tallies should be reset before active batches so they are accumulated without bias. - *Default*: 0 + *Default*: 0 ```` Element ------------------- @@ -1224,9 +1248,9 @@ It can be turned on with "true" and off with "false". ```` Element ---------------------- -The ```` element controls if cmfd tallies should be accumulated -during inactive batches. For some applications, CMFD tallies may not be -needed until the start of active batches. This option can be turned on +The ```` element controls if cmfd tallies should be accumulated +during inactive batches. For some applications, CMFD tallies may not be +needed until the start of active batches. This option can be turned on with "true" and off with "false" *Default*: true @@ -1243,12 +1267,12 @@ occurs. The amout of resets is controlled with the ```` element. ```` Element ------------------------- -The ```` element is used to view the convergence of linear GMRES -iterations in PETSc. This option can be turned on with "true" and turned off +The ```` 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 + *Default*: false ```` Element ------------------ @@ -1261,7 +1285,7 @@ attributes/sub-elements: 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 coordinate are given, it is assumed that the mesh is an x-y mesh. :dimension: @@ -1272,8 +1296,8 @@ attributes/sub-elements: :energy: Energy bins [in MeV], listed in ascending order (e.g. 0.0 0.625e-7 20.0) - for CMFD tallies and acceleration. If no energy bins are listed, OpenMC - automatically assumes a one energy group calculation over the entire + for CMFD tallies and acceleration. If no energy bins are listed, OpenMC + automatically assumes a one energy group calculation over the entire energy range. :albedo: @@ -1283,10 +1307,10 @@ attributes/sub-elements: *Default*: 1.0 1.0 1.0 1.0 1.0 1.0 :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 + 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 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 + For a simple 4x4 coarse mesh with a 2x2 fuel lattice surrounded by reflector, the map is: ``1 1 1 1`` @@ -1297,24 +1321,24 @@ attributes/sub-elements: ``1 1 1 1`` - Therefore a 2x2 system of equations is solved rather than a 4x4. This - is extremely important to use in reflectors as neutrons will not + Therefore a 2x2 system of equations is solved rather than a 4x4. This + is extremely important to use in reflectors as neutrons will not contribute to any tallies far away from fission source neutron regions. A ``2`` must be used to identify any fission source region. - .. note:: Only two of the following three sub-elements are needed: - ``lower_left``, ``upper_right`` and ``width``. Any combination + .. note:: Only two of the following three sub-elements are needed: + ``lower_left``, ``upper_right`` and ``width``. Any combination of two of these will yield the third. ```` Element ------------------ -The ```` element is used to normalize the CMFD fission source distribution -to a particular value. For example, if a fission source is calculated for a -17 x 17 lattice of pins, the fission source may be normalized to the number of -fission source regions, in this case 289. This is useful when visualizing this -distribution as the average peaking factor will be unity. This parameter will -not impact the calculation. +The ```` element is used to normalize the CMFD fission source distribution +to a particular value. For example, if a fission source is calculated for a +17 x 17 lattice of pins, the fission source may be normalized to the number of +fission source regions, in this case 289. This is useful when visualizing this +distribution as the average peaking factor will be unity. This parameter will +not impact the calculation. *Default*: 1.0 @@ -1329,7 +1353,7 @@ occur during inactive CMFD batches. ```` Element --------------------------- -The ```` element is used to view the convergence of power iteration. +The ```` 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 @@ -1355,7 +1379,7 @@ function in PETSc. This option can be turned on with "true" and turned off with -------------------- The ```` element controls whether the CMFD eigenproblem is solved with -standard power iteration or nonlinear Jacobian-free Newton Krylov (JFNK). +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