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Various fixes and updates in documentation
This commit is contained in:
Paul Romano
parent
86c081b28b
commit
5b195b6e9f
10 changed files with 125 additions and 130 deletions
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@ -109,7 +109,8 @@ The current version of the statepoint file format is 17.0.
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**/tallies/tally <uid>/**
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:Attributes: - **internal** (*int*) -- Flag indicating the presence of tally
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:Attributes:
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- **internal** (*int*) -- Flag indicating the presence of tally
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data (0) or absence of tally data (1). All user defined
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tallies will have a value of 0 unless otherwise instructed.
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@ -8,13 +8,14 @@ Cross Section Representations
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Continuous-Energy Data
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----------------------
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The data governing the interaction of neutrons with
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various nuclei for continous-energy problems are represented using the ACE
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format which is used by MCNP_ and Serpent_. ACE-format data can be generated
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with the NJOY_ nuclear data processing system which converts raw
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`ENDF/B data`_ into linearly-interpolable data as required by most Monte Carlo
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codes. The use of a standard cross section format allows for a direct comparison
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of OpenMC with other codes since the same cross section libraries can be used.
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In OpenMC, the data governing the interaction of neutrons with various nuclei
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for continous-energy problems are represented using an HDF5 format that can be
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produced by converting files in the ACE format, which is used by MCNP_ and
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Serpent_. ACE-format data can be generated with the NJOY_ nuclear data
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processing system, which converts raw `ENDF/B data`_ into linearly-interpolable
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data as required by most Monte Carlo codes. Since ACE-format data can be
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converted into OpenMC's HDF5 format, it is possible to perform direct comparison
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of OpenMC with other codes using the same underlying nuclear data library.
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The ACE format contains continuous-energy cross sections for the following types
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of reactions: elastic scattering, fission (or first-chance fission,
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@ -31,7 +32,7 @@ data can be used.
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Energy Grid Methods
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-------------------
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The method by which continuous energy cross sections for each nuclide in a
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The method by which continuous-energy cross sections for each nuclide in a
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problem are stored as a function of energy can have a substantial effect on the
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performance of a Monte Carlo simulation. Since the ACE format is based on
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linearly-interpolable cross sections, each nuclide has cross sections tabulated
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@ -72,9 +73,9 @@ Windowed Multipole Representation
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---------------------------------
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In addition to the usual pointwise representation of cross sections, OpenMC
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offers support for an experimental data format called windowed multipole (WMP).
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This data format requires less memory than pointwise cross sections, and it
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allows on-the-fly Doppler broadening to arbitrary temperature.
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offers support for an data format called windowed multipole (WMP). This data
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format requires less memory than pointwise cross sections, and it allows
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on-the-fly Doppler broadening to arbitrary temperature.
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The multipole method was introduced by Hwang_ and the faster windowed multipole
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method by Josey_. In the multipole format, cross section resonances are
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@ -258,7 +259,7 @@ where a material has a very large cross sections relative to the other material
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used to minimize this error.
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Finally, the above options for representing the physics do not have to be
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consistent across the problem. The number of groups and the structure, however,
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consistent across the problem. The number of groups and the structure, however,
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does have to be consistent across the data sets. That is to say that each
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microscopic or macroscopic data set does not have to apply the same scattering
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expansion, treatment of multiplicity or angular representation of the cross
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@ -4,11 +4,11 @@
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Heating and Energy Deposition
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=============================
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As particles traverse a problem, some portion of their energy is deposited at
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As particles traverse a problem, some portion of their energy is deposited at
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collision sites. This energy is deposited when charged particles, including
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electrons and recoil nuclei, undergo electromagnetic interactions with
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surrounding electons and ions. The information describing how much energy
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is deposited for a specific reaction is referred to as
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is deposited for a specific reaction is referred to as
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"heating numbers" and can be computed using a program like NJOY with the
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``heatr`` module.
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@ -108,7 +108,7 @@ Neutron Transport
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For this case, OpenMC instructs ``heatr`` to produce heating coefficients
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assuming that energy from photons, :math:`E_{\gamma, p}` and
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:math:`E_{\gamma, d}`, is deposited at the fission site.
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Let :math:`N901` represent the total heating number returned from this ``heatr``
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Let :math:`N901` represent the total heating number returned from this ``heatr``
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run with :math:`N918` reflecting fission heating computed from NJOY.
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:math:`M901` represent the following modification
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@ -119,7 +119,7 @@ run with :math:`N918` reflecting fission heating computed from NJOY.
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+ E_{i, \gamma, d}\right]\sigma_{i, f}(E).
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This modified heating data is stored as the MT=901 reaction and will be scored
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if ``901`` is included in :attr:`openmc.Tally.scores`.
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if ``heating-local`` is included in :attr:`openmc.Tally.scores`.
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Coupled neutron-photon transport
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--------------------------------
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@ -146,4 +146,4 @@ References
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.. [Mack97] Abdou, M.A., Maynard, C.W., and Wright, R.Q. MACK: computer
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program to calculate neutron energy release parameters (fluence-to-kerma
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factors) and multigroup neutron reaction cross sections from nuclear data
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in ENDF Format. Oak Ridge National Laboratory report ORNL-TM-3994.
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in ENDF Format. Oak Ridge National Laboratory report ORNL-TM-3994.
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@ -139,7 +139,7 @@ be performed before the run is finished. This include the following:
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- If requested, a source file is written to disk.
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- All allocatable arrays are deallocated.
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- Dynamically-allocated memory should be freed.
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.. _probability distributions: https://en.wikipedia.org/wiki/Probability_distribution
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.. _Monte Carlo: https://en.wikipedia.org/wiki/Monte_Carlo_method
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@ -1298,11 +1298,10 @@ section over the range of velocities considered:
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where it should be noted that the maximum is taken over the range :math:`[v_n -
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4/\beta, 4_n + 4\beta]`. This method is known as Doppler broadening rejection
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correction (DBRC) and was first introduced by `Becker et al.`_. OpenMC has an
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implementation of DBRC as well as an accelerated sampling method that are
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described fully in `Walsh et al.`_
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implementation of DBRC as well as an accelerated sampling method that samples the `relative velocity`_ directly.
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.. _Becker et al.: https://doi.org/10.1016/j.anucene.2008.12.001
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.. _Walsh et al.: https://doi.org/10.1016/j.anucene.2014.01.017
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.. _relative velocity: https://doi.org/10.1016/j.anucene.2017.12.044
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.. _sab_tables:
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@ -2,6 +2,8 @@
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:mod:`openmc.data` -- Nuclear Data Interface
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--------------------------------------------
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.. module:: openmc.data
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Core Classes
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------------
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@ -13,15 +15,15 @@ and product yields.
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:nosignatures:
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:template: myclass.rst
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openmc.data.IncidentNeutron
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openmc.data.Reaction
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openmc.data.Product
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openmc.data.FissionEnergyRelease
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openmc.data.DataLibrary
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openmc.data.Decay
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openmc.data.FissionProductYields
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openmc.data.WindowedMultipole
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openmc.data.ProbabilityTables
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IncidentNeutron
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Reaction
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Product
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FissionEnergyRelease
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DataLibrary
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Decay
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FissionProductYields
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WindowedMultipole
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ProbabilityTables
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The following classes are used for storing atomic data (incident photon cross
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sections, atomic relaxation):
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@ -31,9 +33,9 @@ sections, atomic relaxation):
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:nosignatures:
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:template: myclass.rst
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openmc.data.IncidentPhoton
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openmc.data.PhotonReaction
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openmc.data.AtomicRelaxation
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IncidentPhoton
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PhotonReaction
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AtomicRelaxation
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The following classes are used for storing thermal neutron scattering data:
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@ -43,10 +45,10 @@ The following classes are used for storing thermal neutron scattering data:
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:nosignatures:
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:template: myclass.rst
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openmc.data.ThermalScattering
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openmc.data.ThermalScatteringReaction
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openmc.data.CoherentElastic
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openmc.data.IncoherentElastic
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ThermalScattering
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ThermalScatteringReaction
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CoherentElastic
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IncoherentElastic
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Core Functions
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@ -57,12 +59,12 @@ Core Functions
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:nosignatures:
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:template: myfunction.rst
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openmc.data.atomic_mass
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openmc.data.gnd_name
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openmc.data.linearize
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openmc.data.thin
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openmc.data.water_density
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openmc.data.zam
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atomic_mass
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gnd_name
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linearize
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thin
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water_density
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zam
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One-dimensional Functions
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-------------------------
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@ -72,13 +74,13 @@ One-dimensional Functions
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:nosignatures:
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:template: myclass.rst
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openmc.data.Function1D
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openmc.data.Tabulated1D
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openmc.data.Polynomial
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openmc.data.Combination
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openmc.data.Sum
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openmc.data.Regions1D
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openmc.data.ResonancesWithBackground
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Function1D
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Tabulated1D
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Polynomial
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Combination
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Sum
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Regions1D
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ResonancesWithBackground
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Angle-Energy Distributions
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--------------------------
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@ -88,27 +90,27 @@ Angle-Energy Distributions
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:nosignatures:
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:template: myclass.rst
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openmc.data.AngleEnergy
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openmc.data.KalbachMann
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openmc.data.CorrelatedAngleEnergy
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openmc.data.UncorrelatedAngleEnergy
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openmc.data.NBodyPhaseSpace
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openmc.data.LaboratoryAngleEnergy
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openmc.data.AngleDistribution
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openmc.data.EnergyDistribution
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openmc.data.ArbitraryTabulated
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openmc.data.GeneralEvaporation
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openmc.data.MaxwellEnergy
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openmc.data.Evaporation
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openmc.data.WattEnergy
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openmc.data.MadlandNix
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openmc.data.DiscretePhoton
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openmc.data.LevelInelastic
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openmc.data.ContinuousTabular
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openmc.data.CoherentElasticAE
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openmc.data.IncoherentElasticAE
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openmc.data.IncoherentElasticAEDiscrete
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openmc.data.IncoherentInelasticAEDiscrete
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AngleEnergy
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KalbachMann
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CorrelatedAngleEnergy
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UncorrelatedAngleEnergy
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NBodyPhaseSpace
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LaboratoryAngleEnergy
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AngleDistribution
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EnergyDistribution
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ArbitraryTabulated
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GeneralEvaporation
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MaxwellEnergy
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Evaporation
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WattEnergy
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MadlandNix
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DiscretePhoton
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LevelInelastic
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ContinuousTabular
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CoherentElasticAE
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IncoherentElasticAE
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IncoherentElasticAEDiscrete
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IncoherentInelasticAEDiscrete
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Resonance Data
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--------------
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@ -118,20 +120,20 @@ Resonance Data
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:nosignatures:
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:template: myclass.rst
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openmc.data.Resonances
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openmc.data.ResonanceRange
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openmc.data.SingleLevelBreitWigner
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openmc.data.MultiLevelBreitWigner
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openmc.data.ReichMoore
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openmc.data.RMatrixLimited
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openmc.data.ResonanceCovariances
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openmc.data.ResonanceCovarianceRange
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openmc.data.SingleLevelBreitWignerCovariance
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openmc.data.MultiLevelBreitWignerCovariance
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openmc.data.ReichMooreCovariance
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openmc.data.ParticlePair
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openmc.data.SpinGroup
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openmc.data.Unresolved
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Resonances
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ResonanceRange
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SingleLevelBreitWigner
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MultiLevelBreitWigner
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ReichMoore
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RMatrixLimited
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ResonanceCovariances
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ResonanceCovarianceRange
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SingleLevelBreitWignerCovariance
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MultiLevelBreitWignerCovariance
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ReichMooreCovariance
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ParticlePair
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SpinGroup
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Unresolved
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ACE Format
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----------
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@ -144,8 +146,8 @@ Classes
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:nosignatures:
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:template: myclass.rst
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openmc.data.ace.Library
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openmc.data.ace.Table
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ace.Library
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ace.Table
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Functions
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+++++++++
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@ -155,7 +157,7 @@ Functions
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:nosignatures:
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:template: myfunction.rst
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openmc.data.ace.ascii_to_binary
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ace.ascii_to_binary
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ENDF Format
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-----------
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@ -168,7 +170,7 @@ Classes
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:nosignatures:
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:template: myclass.rst
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openmc.data.endf.Evaluation
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endf.Evaluation
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Functions
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+++++++++
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@ -178,13 +180,13 @@ Functions
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:nosignatures:
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:template: myfunction.rst
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openmc.data.endf.float_endf
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openmc.data.endf.get_cont_record
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openmc.data.endf.get_evaluations
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openmc.data.endf.get_head_record
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openmc.data.endf.get_tab1_record
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openmc.data.endf.get_tab2_record
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openmc.data.endf.get_text_record
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endf.float_endf
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endf.get_cont_record
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endf.get_evaluations
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endf.get_head_record
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endf.get_tab1_record
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endf.get_tab2_record
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endf.get_text_record
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NJOY Interface
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--------------
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@ -194,7 +196,7 @@ NJOY Interface
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:nosignatures:
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:template: myfunction.rst
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openmc.data.njoy.run
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openmc.data.njoy.make_pendf
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openmc.data.njoy.make_ace
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openmc.data.njoy.make_ace_thermal
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njoy.run
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njoy.make_pendf
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njoy.make_ace
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njoy.make_ace_thermal
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@ -1,11 +1,11 @@
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.. _pythonapi_deplete:
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.. module:: openmc.deplete
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----------------------------------
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:mod:`openmc.deplete` -- Depletion
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----------------------------------
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.. module:: openmc.deplete
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Primary API
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-----------
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@ -16,10 +16,10 @@ recommended to use one of the pregenerated libraries. Alternatively, if you have
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ACE format data that was produced with NJOY_, such as that distributed with
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MCNP_ or Serpent_, it can be converted to the HDF5 format using the :ref:`using
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the Python API <create_xs_library>`. Several sources provide openly available
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ACE data including the `ENDF/B`_, JEFF_, and TENDL_
|
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libraries. In addition to tabulated cross sections in the HDF5 files, OpenMC
|
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relies on :ref:`windowed multipole <windowed_multipole>` data to perform
|
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on-the-fly Doppler broadening.
|
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ACE data including the `ENDF/B`_, JEFF_, and TENDL_ libraries as well as the
|
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`LANL Nuclear Data Team <https://nucleardata.lanl.gov/>`_. In addition to
|
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tabulated cross sections in the HDF5 files, OpenMC relies on :ref:`windowed
|
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multipole <windowed_multipole>` data to perform on-the-fly Doppler broadening.
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In multi-group mode, OpenMC utilizes an HDF5-based library format which can be
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used to describe nuclide- or material-specific quantities.
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|
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@ -30,11 +30,11 @@ Environment Variables
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When :ref:`scripts_openmc` is run, it will look for several environment
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variables that indicate where cross sections can be found. While the location of
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cross sections can also be indicated through the :class:`openmc.Materials` class
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(or in the :ref:`materials.xml <io_materials>` file), if you always use the same
|
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set of cross section data, it is often easier to just set an environment
|
||||
variable that will be picked up by default every time OpenMC is run. The
|
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following environment variables are used:
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cross sections can also be indicated through the
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:attr:`openmc.Materials.cross_setion` attribute (or in the :ref:`materials.xml
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<io_materials>` file), if you always use the same set of cross section data, it
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is often easier to just set an environment variable that will be picked up by
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default every time OpenMC is run. The following environment variables are used:
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:envvar:`OPENMC_CROSS_SECTIONS`
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Indicates the path to the :ref:`cross_sections.xml <io_cross_sections>`
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|
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@ -121,11 +121,11 @@ For many regions, a bounding-box can be determined automatically::
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While a bounding box can be determined for regions involving half-spaces of
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spheres, cylinders, and axis-aligned planes, it generally cannot be determined
|
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if the region involves cones, non-axis-aligned planes, or other exotic
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second-order surfaces. For example, the :func:`openmc.get_hexagonal_prism`
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second-order surfaces. For example, the :func:`openmc.model.hexagonal_prism`
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function returns the interior region of a hexagonal prism; because it is bounded
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by a :class:`openmc.Plane`, trying to get its bounding box won't work::
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>>> hex = openmc.get_hexagonal_prism()
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>>> hex = openmc.model.hexagonal_prism()
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>>> hex.bounding_box
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(array([-0.8660254, -inf, -inf]),
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array([ 0.8660254, inf, inf]))
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@ -374,7 +374,7 @@ code would work::
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hexlat.universes = [outer_ring, middle_ring, inner_ring]
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If you need to create a hexagonal boundary (composed of six planar surfaces) for
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a hexagonal lattice, :func:`openmc.get_hexagonal_prism` can be used.
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a hexagonal lattice, :func:`openmc.model.hexagonal_prism` can be used.
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|
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.. _usersguide_geom_export:
|
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@ -396,6 +396,13 @@ if needed, lattices, the last step is to create an instance of
|
|||
geom.root_universe = root_univ
|
||||
geom.export_to_xml()
|
||||
|
||||
Note that it's not strictly required to manually create a root universe. You can
|
||||
also pass a list of cells to the :class:`openmc.Geometry` constructor and it
|
||||
will handle creating the unverse::
|
||||
|
||||
geom = openmc.Geometry([cell1, cell2, cell3])
|
||||
geom.export_to_xml()
|
||||
|
||||
.. _constructive solid geometry: https://en.wikipedia.org/wiki/Constructive_solid_geometry
|
||||
.. _quadratic surfaces: https://en.wikipedia.org/wiki/Quadric
|
||||
|
||||
|
|
|
|||
|
|
@ -221,26 +221,11 @@ selected::
|
|||
|
||||
settings.electron_treatment = 'led'
|
||||
|
||||
.. warning::
|
||||
Currently, collision stopping powers used in the TTB approximation come from
|
||||
the `NIST ESTAR database`_, which provides data for each element calculated
|
||||
using by default the material density at standard temperature and pressure.
|
||||
In OpenMC, stopping powers for compounds are calculated from this elemental
|
||||
data using Bragg's additivity rule. However, this is not a good
|
||||
approximation --- the collision stopping power is a function of certain
|
||||
quantities, such as the mean excitation energy and particularly the density
|
||||
effect correction, that depend on material properties. Data for constituent
|
||||
elements in a compound cannot simply be summed together, but rather these
|
||||
quantities should be calculated for the material. This treatment will be
|
||||
especially poor when the density of a material is different from the
|
||||
densities used in the NIST data.
|
||||
|
||||
.. note::
|
||||
Some features related to photon transport are not currently implemented,
|
||||
including:
|
||||
|
||||
* Tallying photon energy deposition.
|
||||
* Properly accounting for energy deposition in coupled n-p calculations.
|
||||
* Generating a photon source from a neutron calculation that can be used
|
||||
for a later fixed source photon calculation.
|
||||
* Photoneutron reactions.
|
||||
|
|
|
|||
Loading…
Add table
Add a link
Reference in a new issue