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Fixed merge conflicts with develop
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Will Boyd
commit
4e6aca6028
331 changed files with 6506 additions and 222815 deletions
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@ -441,24 +441,22 @@ attributes/sub-elements:
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has the following attributes:
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:type:
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The type of spatial distribution. Valid options are "box" and "point". A
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"box" spatial distribution has coordinates sampled uniformly in a
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parallelepiped. A "point" spatial distribution has coordinates specified
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by a triplet.
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The type of spatial distribution. Valid options are "box", "fission", and
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"point". A "box" spatial distribution has coordinates sampled uniformly in
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a parallelepiped. A "fission" spatial distribution samples locations from
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a "box" distribution but only locations in fissionable materials are
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accepted. A "point" spatial distribution has coordinates specified by a
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triplet.
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*Default*: None
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:parameters:
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For a "box" spatial distribution, ``parameters`` should be given as six
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real numbers, the first three of which specify the lower-left corner of a
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parallelepiped and the last three of which specify the upper-right
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corner. Source sites are sampled uniformly through that parallelepiped.
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To filter a "box" spatial distribution by fissionable material, specify
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"fission" tag instead of "box". The ``parameters`` should be given as six
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real numbers, the first three of which specify the lower-left corner of a
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parallelepiped and the last three of which specify the upper-right
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corner. Source sites are sampled uniformly through that parallelepiped.
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For a "box" or "fission" spatial distribution, ``parameters`` should be
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given as six real numbers, the first three of which specify the lower-left
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corner of a parallelepiped and the last three of which specify the
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upper-right corner. Source sites are sampled uniformly through that
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parallelepiped.
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For a "point" spatial distribution, ``parameters`` should be given as
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three real numbers which specify the (x,y,z) location of an isotropic
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@ -481,7 +479,7 @@ attributes/sub-elements:
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:parameters:
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For an "isotropic" angular distribution, ``parameters`` should not be
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specified
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specified.
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For a "monodirectional" angular distribution, ``parameters`` should be
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given as three real numbers which specify the angular cosines with respect
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@ -499,7 +497,7 @@ attributes/sub-elements:
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"watt", and "maxwell". The "monoenergetic" option produces source sites at
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a single energy. The "watt" option produces source sites whose energy is
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sampled from a Watt fission spectrum. The "maxwell" option produce source
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sites whose energy is sampled from a Maxwell fission spectrum
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sites whose energy is sampled from a Maxwell fission spectrum.
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*Default*: watt
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@ -605,8 +603,6 @@ survival biasing, otherwise known as implicit capture or absorption.
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*Default*: false
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.. _trace:
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``<threads>`` Element
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---------------------
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@ -615,6 +611,8 @@ a simulation. It has no attributes and accepts a positive integer value.
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*Default*: None (Determined by environment variable :envvar:`OMP_NUM_THREADS`)
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.. _trace:
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``<trace>`` Element
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-------------------
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@ -631,9 +629,9 @@ integers: the batch number, generation number, and particle number.
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The ``<track>`` element specifies particles for which OpenMC will output binary
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files describing particle position at every step of its transport. This element
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should be followed by triplets of integers. Each triplet describes one particle
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. The integers in each triplet specify the batch number, generation number, and
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particle number, respectively.
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should be followed by triplets of integers. Each triplet describes one
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particle. The integers in each triplet specify the batch number, generation
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number, and particle number, respectively.
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*Default*: None
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@ -1224,8 +1222,8 @@ The ``<tally>`` element accepts the following sub-elements:
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The ``filter`` element has the following attributes/sub-elements:
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:type:
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The type of the filter. Accepted options are "cell", "cellborn",
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"material", "universe", "energy", "energyout", "mesh", and
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The type of the filter. Accepted options are "cell", "cellborn",
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"material", "universe", "energy", "energyout", "mesh", and
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"distribcell".
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:bins:
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@ -1307,24 +1305,25 @@ The ``<tally>`` element accepts the following sub-elements:
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physical quantities:
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:flux:
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Total flux
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Total flux in particle-cm per source particle.
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:total:
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Total reaction rate
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Total reaction rate in reactions per source particle.
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:scatter:
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Total scattering rate. Can also be identified with the ``scatter-0``
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response type.
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response type. Units are reactions per source particle.
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:absorption:
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Total absorption rate. This accounts for all reactions which do not
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produce secondary neutrons.
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produce secondary neutrons. Units are reactions per source particle.
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:fission:
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Total fission rate
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Total fission rate in reactions per source particle.
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:nu-fission:
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Total production of neutrons due to fission
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Total production of neutrons due to fission. Units are neutrons produced
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per source neutron.
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:kappa-fission:
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The recoverable energy production rate due to fission. The recoverable
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@ -1333,51 +1332,55 @@ The ``<tally>`` element accepts the following sub-elements:
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total energies, and the total energy released by the delayed :math:`\beta`
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particles. The neutrino energy does not contribute to this response. The
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prompt and delayed :math:`\gamma`-rays are assumed to deposit their energy
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locally.
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locally. Units are MeV per source particle.
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:scatter-N:
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Tally the N\ :sup:`th` \ scattering moment, where N is the Legendre
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expansion order of the change in particle angle :math:`\left(\mu\right)`.
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N must be between 0 and 10. As an example, tallying the
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2\ :sup:`nd` \ scattering moment would be specified as
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``<scores> scatter-2 </scores>``.
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N must be between 0 and 10. As an example, tallying the 2\ :sup:`nd` \
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scattering moment would be specified as ``<scores> scatter-2
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</scores>``. Units are reactions per source particle.
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:scatter-PN:
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Tally all of the scattering moments from order 0 to N, where N is the
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Legendre expansion order of the change in particle angle
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:math:`\left(\mu\right)`. That is, ``scatter-P1`` is equivalent to
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requesting tallies of ``scatter-0`` and ``scatter-1``. Like for
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``scatter-N``, N must be between 0 and 10. As an example, tallying up
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to the 2\ :sup:`nd` \ scattering moment would be specified as
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``<scores> scatter-P2 </scores>``.
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``scatter-N``, N must be between 0 and 10. As an example, tallying up to
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the 2\ :sup:`nd` \ scattering moment would be specified as ``<scores>
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scatter-P2 </scores>``. Units are reactions per source particle.
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:scatter-YN:
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``scatter-YN`` is similar to ``scatter-PN`` except an additional
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expansion is performed for the incoming particle direction
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``scatter-YN`` is similar to ``scatter-PN`` except an additional expansion
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is performed for the incoming particle direction
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:math:`\left(\Omega\right)` using the real spherical harmonics. This is
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useful for performing angular flux moment weighting of the scattering
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moments. Like ``scatter-PN``, ``scatter-YN`` will tally all of the
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moments from order 0 to N; N again must be between 0 and 10.
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moments. Like ``scatter-PN``, ``scatter-YN`` will tally all of the moments
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from order 0 to N; N again must be between 0 and 10. Units are reactions
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per source particle.
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:nu-scatter, nu-scatter-N, nu-scatter-PN, nu-scatter-YN:
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These scores are similar in functionality to their ``scatter*``
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equivalents except the total production of neutrons due to
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scattering is scored vice simply the scattering rate. This accounts for
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multiplicity from (n,2n), (n,3n), and (n,4n) reactions.
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equivalents except the total production of neutrons due to scattering is
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scored vice simply the scattering rate. This accounts for multiplicity
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from (n,2n), (n,3n), and (n,4n) reactions. Units are neutrons produced per
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source particle.
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:flux-YN:
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Spherical harmonic expansion of the direction of motion
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:math:`\left(\Omega\right)` of the total flux. This score will tally
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all of the harmonic moments of order 0 to N. N must be between 0 and 10.
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:math:`\left(\Omega\right)` of the total flux. This score will tally all
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of the harmonic moments of order 0 to N. N must be between 0
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and 10. Units are particle-cm per source particle.
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:total-YN:
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The total reaction rate expanded via spherical harmonics about the
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direction of motion of the neutron, :math:`\Omega`.
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This score will tally all of the harmonic moments of order 0 to N. N must
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be between 0 and 10.
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be between 0 and 10. Units are reactions per source particle.
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:current:
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Partial currents on the boundaries of each cell in a mesh.
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Partial currents on the boundaries of each cell in a mesh. Units are
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particles per source particle.
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.. note::
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This score can only be used if a mesh filter has been
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@ -1385,7 +1388,7 @@ The ``<tally>`` element accepts the following sub-elements:
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other score.
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:events:
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Number of scoring events
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Number of scoring events. Units are events per source particle.
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:trigger:
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Precision trigger applied to all filter bins and nuclides for this tally.
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@ -1675,6 +1678,15 @@ The ``<begin>`` element controls what batch CMFD calculations should begin.
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*Default*: 1
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``<dhat_reset>`` Element
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------------------------
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The ``<dhat_reset>`` element controls whether :math:`\widehat{D}` nonlinear
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CMFD parameters should be reset to zero before solving CMFD eigenproblem.
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It can be turned on with "true" and off with "false".
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*Default*: false
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``<display>`` Element
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---------------------
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@ -1691,15 +1703,6 @@ The ``<display>`` element sets one additional CMFD output column. Options are:
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*Default*: balance
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``<dhat_reset>`` Element
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------------------------
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The ``<dhat_reset>`` element controls whether :math:`\widehat{D}` nonlinear
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CMFD parameters should be reset to zero before solving CMFD eigenproblem.
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It can be turned on with "true" and off with "false".
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*Default*: false
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``<downscatter>`` Element
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-------------------------
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@ -1743,11 +1746,11 @@ The CMFD mesh is a structured Cartesian mesh. This element has the following
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attributes/sub-elements:
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:lower_left:
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The lower-left corner of the structured mesh. If only two coordinate are
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The lower-left corner of the structured mesh. If only two coordinates are
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given, it is assumed that the mesh is an x-y mesh.
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:upper_right:
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The upper-right corner of the structrued mesh. If only two coordinate are
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The upper-right corner of the structrued mesh. If only two coordinates are
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given, it is assumed that the mesh is an x-y mesh.
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:dimension:
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@ -1770,7 +1773,7 @@ attributes/sub-elements:
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:map:
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An optional acceleration map can be specified to overlay on the coarse
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mesh spatial grid. If this option is used a ``1`` is used for a
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mesh spatial grid. If this option is used, a ``1`` is used for a
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non-accelerated region and a ``2`` is used for an accelerated region.
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For a simple 4x4 coarse mesh with a 2x2 fuel lattice surrounded by
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reflector, the map is:
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@ -160,6 +160,13 @@ openmp
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Enables shared-memory parallelism using the OpenMP API. The Fortran compiler
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being used must support OpenMP.
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coverage
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Compile and link code instrumented for coverage analysis. This is typically
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used in conjunction with gcov_.
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maxcoord
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Maximum number of nested coordinate levels in geometry. Defaults to 10.
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To set any of these options (e.g. turning on debug mode), the following form
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should be used:
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@ -167,6 +174,8 @@ should be used:
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cmake -Ddebug=on /path/to/openmc
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.. _gcov: https://gcc.gnu.org/onlinedocs/gcc/Gcov.html
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Compiling with MPI
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++++++++++++++++++
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