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Merge branch 'develop' into diff_tally6
This commit is contained in:
Sterling Harper
commit
bbcc4f6453
367 changed files with 44247 additions and 39004 deletions
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@ -8,3 +8,15 @@
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|||
max-width: 100%;
|
||||
overflow: visible;
|
||||
}
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||||
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||||
.wy-plain-list-disc, .rst-content .section ul, .rst-content .toctree-wrapper ul, article ul {
|
||||
margin-bottom: 0px;
|
||||
}
|
||||
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||||
.wy-table, .rst-content table.docutils, .rst-content table.field-list {
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||||
margin-bottom: 0px;
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}
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||||
.wy-side-nav-search {
|
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background-color: #343131;
|
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}
|
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|
|
|
|||
8
docs/source/_templates/myclassinherit.rst
Normal file
8
docs/source/_templates/myclassinherit.rst
Normal file
|
|
@ -0,0 +1,8 @@
|
|||
{{ fullname }}
|
||||
{{ underline }}
|
||||
|
||||
.. currentmodule:: {{ module }}
|
||||
|
||||
.. autoclass:: {{ objname }}
|
||||
:members:
|
||||
:inherited-members:
|
||||
|
|
@ -24,9 +24,13 @@ except ImportError:
|
|||
from mock import Mock as MagicMock
|
||||
|
||||
|
||||
MOCK_MODULES = ['numpy', 'h5py', 'pandas', 'opencg']
|
||||
MOCK_MODULES = ['numpy', 'numpy.polynomial', 'numpy.polynomial.polynomial',
|
||||
'h5py', 'pandas', 'opencg']
|
||||
sys.modules.update((mod_name, MagicMock()) for mod_name in MOCK_MODULES)
|
||||
|
||||
import numpy as np
|
||||
np.polynomial.Polynomial = MagicMock
|
||||
|
||||
|
||||
# If extensions (or modules to document with autodoc) are in another directory,
|
||||
# add these directories to sys.path here. If the directory is relative to the
|
||||
|
|
@ -69,9 +73,9 @@ copyright = u'2011-2016, Massachusetts Institute of Technology'
|
|||
# built documents.
|
||||
#
|
||||
# The short X.Y version.
|
||||
version = "0.7"
|
||||
version = "0.8"
|
||||
# The full version, including alpha/beta/rc tags.
|
||||
release = "0.7.1"
|
||||
release = "0.8.0"
|
||||
|
||||
# The language for content autogenerated by Sphinx. Refer to documentation
|
||||
# for a list of supported languages.
|
||||
|
|
@ -125,7 +129,7 @@ if not on_rtd:
|
|||
html_theme = 'sphinx_rtd_theme'
|
||||
html_theme_path = [sphinx_rtd_theme.get_html_theme_path()]
|
||||
|
||||
html_logo = '_images/openmc200px.png'
|
||||
html_logo = '_images/openmc_logo.png'
|
||||
|
||||
# The name for this set of Sphinx documents. If None, it defaults to
|
||||
# "<project> v<release> documentation".
|
||||
|
|
|
|||
|
|
@ -5,9 +5,9 @@ The OpenMC Monte Carlo Code
|
|||
OpenMC is a Monte Carlo particle transport simulation code focused on neutron
|
||||
criticality calculations. It is capable of simulating 3D models based on
|
||||
constructive solid geometry with second-order surfaces. OpenMC supports either
|
||||
continuous-energy or multi-group transport. The continuous-energy
|
||||
particle interaction data is based on ACE format cross sections, also used
|
||||
in the MCNP and Serpent Monte Carlo codes.
|
||||
continuous-energy or multi-group transport. The continuous-energy particle
|
||||
interaction data is based on a native HDF5 format that can be generated from ACE
|
||||
files used by the MCNP and Serpent Monte Carlo codes.
|
||||
|
||||
OpenMC was originally developed by members of the `Computational Reactor Physics
|
||||
Group`_ at the `Massachusetts Institute of Technology`_ starting
|
||||
|
|
|
|||
|
|
@ -1,8 +1,11 @@
|
|||
.. _io_data_wmp:
|
||||
|
||||
==========================================
|
||||
The Windowed Multipole Library Format v0.2
|
||||
==========================================
|
||||
=================================
|
||||
Windowed Multipole Library Format
|
||||
=================================
|
||||
|
||||
**/version** (*char[]*)
|
||||
The format version of the file. The current version is "v0.2"
|
||||
|
||||
**/nuclide/**
|
||||
- **broaden_poly** (*int[]*)
|
||||
|
|
@ -12,14 +15,17 @@ The Windowed Multipole Library Format v0.2
|
|||
Curve fit coefficients. Indexed by (reaction type, coefficient index,
|
||||
window index).
|
||||
- **data** (*complex[][]*)
|
||||
Complex poles and residues. Each pole has a corresponding set of
|
||||
residues. For example, the `i`th pole and corresponding residues are
|
||||
stored as `data[:,i] = [pole, residue_1, residue_2, ...]`. The
|
||||
residues are in the order: total, competitive if present, absorption,
|
||||
fission. Complex numbers are stored by forming a type with `"r"` and
|
||||
`"i"` identifiers, similar to how `h5py` does it.
|
||||
- **start_E** (*double*)
|
||||
Lowest energy the windowed multipole part of the library is valid for.
|
||||
Complex poles and residues. Each pole has a corresponding set of
|
||||
residues. For example, the :math:`i`-th pole and corresponding residues
|
||||
are stored as
|
||||
|
||||
.. math::
|
||||
\text{data}[:,i] = [\text{pole},~\text{residue}_1,~\text{residue}_2,
|
||||
~\ldots]
|
||||
|
||||
The residues are in the order: total, competitive if present,
|
||||
absorption, fission. Complex numbers are stored by forming a type with
|
||||
":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[]*)
|
||||
|
|
@ -59,24 +65,27 @@ The Windowed Multipole Library Format v0.2
|
|||
Number of possible :math:`l` quantum states for this nuclide.
|
||||
- **pseudo_K0RS** (*double[]*)
|
||||
:math:`l` dependent value of
|
||||
|
||||
|
||||
.. math::
|
||||
\sqrt{\frac{2 m_n}{\hbar}}\frac{AWR}{AWR + 1} r_{s,l}
|
||||
|
||||
|
||||
Where :math:`m_n` is mass of neutron, :math:`AWR` is the atomic weight
|
||||
ratio of the target to the neutron, and :math:`r_{s,l}` is the
|
||||
scattering radius for a given :math:`l`.
|
||||
- **spacing** (*double*)
|
||||
.. math::
|
||||
\frac{\sqrt{E_{max}}- \sqrt{E_{min}}}{n_w}
|
||||
|
||||
|
||||
Where :math:`E_{max}` is the maximum energy the windows go up to. This
|
||||
is not equivalent to the maximum energy for which the windowed multipole
|
||||
data is valid for. It is slightly higher to ensure an integer number of
|
||||
windows. :math:`E_{min}` is the minimum energy and equivalent to
|
||||
`start_E`, and :math:`n_w` is the number of windows, given by `windows`.
|
||||
``start_E``, and :math:`n_w` is the number of windows, given by
|
||||
``windows``.
|
||||
- **sqrtAWR** (*double*)
|
||||
Square root of the atomic weight ratio.
|
||||
- **start_E** (*double*)
|
||||
Lowest energy the windowed multipole part of the library is valid for.
|
||||
- **w_start** (*int[]*)
|
||||
The pole to start from for each window.
|
||||
- **w_end** (*int[]*)
|
||||
|
|
@ -87,6 +96,7 @@ The Windowed Multipole Library Format v0.2
|
|||
**/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`.
|
||||
- **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
|
||||
|
|
|
|||
53
docs/source/io_formats/fission_energy.rst
Normal file
53
docs/source/io_formats/fission_energy.rst
Normal file
|
|
@ -0,0 +1,53 @@
|
|||
.. _usersguide_fission_energy:
|
||||
|
||||
==================================
|
||||
Fission Energy Release File Format
|
||||
==================================
|
||||
|
||||
This file is a compact HDF5 representation of the ENDF MT=1, MF=458 data (see
|
||||
ENDF-102_ for details). It gives the information needed to compute the energy
|
||||
carried away from fission reactions by each reaction product (e.g. fragment
|
||||
nuclei, neutrons) which depends on the incident neutron energy. OpenMC is
|
||||
distributed with one of these files under
|
||||
data/fission_Q_data_endfb71.h5. More files of this format can be created from
|
||||
ENDF files with the
|
||||
``openmc.data.write_compact_458_library`` function. They can be read with the
|
||||
``openmc.data.FissionEnergyRelease.from_compact_hdf5`` class method.
|
||||
|
||||
:Attributes: - **comment** (*char[]*) -- An optional text comment
|
||||
- **component order** (*char[][]*) -- An array of strings
|
||||
specifying the order each reaction product occurs in the data
|
||||
arrays. The components use the 2-3 letter abbreviations
|
||||
specified in ENDF-102 e.g. EFR for fission fragments and ENP for
|
||||
prompt neutrons.
|
||||
|
||||
**/<nuclide name>/**
|
||||
Nuclides are named by concatenating their atomic symbol and mass number. For
|
||||
example, 'U235' or 'Pu239'. Metastable nuclides are appended with an
|
||||
'_m' and their metastable number. For example, 'Am242_m1'
|
||||
|
||||
:Datasets:
|
||||
- **data** (*double[][][]*) -- The energy release coefficients. The
|
||||
first axis indexes the component type. The second axis specifies
|
||||
values or uncertainties. The third axis indexes the polynomial
|
||||
order. If the data uses the Sher-Beck format, then the last axis
|
||||
will have a length of one and ENDF-102 should be consulted for
|
||||
energy dependence. Otherwise, the data uses the Madland format
|
||||
which is a polynomial of incident energy.
|
||||
|
||||
For example, if 'EFR' is given first in the **component order**
|
||||
attribute and the data uses the Madland format, then the energy
|
||||
released in the form of fission fragments at an incident energy
|
||||
:math:`E` is given by
|
||||
|
||||
.. math::
|
||||
\text{data}[0, 0, 0] + \text{data}[0, 0, 1] \cdot E
|
||||
+ \text{data}[0, 0, 2] \cdot E^2 + \ldots
|
||||
|
||||
And its uncertainty is
|
||||
|
||||
.. math::
|
||||
\text{data}[0, 1, 0] + \text{data}[0, 1, 1] \cdot E
|
||||
+ \text{data}[0, 1, 2] \cdot E^2 + \ldots
|
||||
|
||||
.. _ENDF-102: http://www.nndc.bnl.gov/endfdocs/ENDF-102-2012.pdf
|
||||
|
|
@ -1,18 +1,34 @@
|
|||
.. _io_file_formats:
|
||||
|
||||
===============
|
||||
IO File Formats
|
||||
===============
|
||||
==========================
|
||||
File Format Specifications
|
||||
==========================
|
||||
|
||||
----------
|
||||
Data Files
|
||||
----------
|
||||
|
||||
.. toctree::
|
||||
:numbered:
|
||||
:maxdepth: 3
|
||||
:maxdepth: 2
|
||||
|
||||
data_wmp
|
||||
nuclear_data
|
||||
mgxs_library
|
||||
data_wmp
|
||||
fission_energy
|
||||
|
||||
------------
|
||||
Output Files
|
||||
------------
|
||||
|
||||
.. toctree::
|
||||
:numbered:
|
||||
:maxdepth: 2
|
||||
|
||||
statepoint
|
||||
source
|
||||
summary
|
||||
particle_restart
|
||||
track
|
||||
voxel
|
||||
volume
|
||||
|
|
|
|||
|
|
@ -22,9 +22,11 @@ materials.
|
|||
|
||||
.. _XML: http://www.w3.org/XML/
|
||||
|
||||
--------------------------------------
|
||||
MGXS Library Specification -- mgxs.xml
|
||||
--------------------------------------
|
||||
.. _mgxs_lib_spec:
|
||||
|
||||
--------------------------
|
||||
MGXS Library Specification
|
||||
--------------------------
|
||||
|
||||
The multi-group library meta-data is contained within the groups_,
|
||||
group_structure_, and inverse_velocities_ elements.
|
||||
|
|
@ -171,9 +173,9 @@ attributes/sub-elements required to describe the meta-data:
|
|||
provided via the ``scatt_type`` element above, is represented and thus used
|
||||
during the scattering process. Specifically, the options are to either
|
||||
convert the Legendre expansion to a tabular representation or leave it as
|
||||
a set of Legendre coefficients. Converting to a tabular representation will
|
||||
cost memory but is likely to decrease runtime compared to leaving as a
|
||||
set of Legendre coefficients. This element has the following
|
||||
a set of Legendre coefficients. Converting to a tabular representation
|
||||
will cost memory but can allow for a decrease in runtime compared to
|
||||
leaving as a set of Legendre coefficients. This element has the following
|
||||
attributes/sub-elements:
|
||||
|
||||
:enable:
|
||||
|
|
|
|||
426
docs/source/io_formats/nuclear_data.rst
Normal file
426
docs/source/io_formats/nuclear_data.rst
Normal file
|
|
@ -0,0 +1,426 @@
|
|||
.. _io_nuclear_data:
|
||||
|
||||
========================
|
||||
Nuclear Data File Format
|
||||
========================
|
||||
|
||||
---------------------
|
||||
Incident Neutron Data
|
||||
---------------------
|
||||
|
||||
|
||||
**/<nuclide name>/**
|
||||
|
||||
:Attributes: - **Z** (*int*) -- Atomic number
|
||||
- **A** (*int*) -- Mass number. For a natural element, A=0 is given.
|
||||
- **metastable** (*int*) -- Metastable state (0=ground, 1=first
|
||||
excited, etc.)
|
||||
- **atomic_weight_ratio** (*double*) -- Mass in units of neutron masses
|
||||
- **n_reaction** (*int*) -- Number of reactions
|
||||
|
||||
:Datasets: - **energy** (*double[]*) -- Energy points at which cross sections are tabulated
|
||||
|
||||
**/<nuclide name>/kTs/**
|
||||
|
||||
<TTT>K is the temperature in Kelvin, rounded to the nearest integer, of the
|
||||
temperature-dependent data set. For example, the data set corresponding to
|
||||
300 Kelvin would be located at `300K`.
|
||||
|
||||
:Datasets:
|
||||
- **<TTT>K** (*double*) -- kT values (in MeV) for each Temperature
|
||||
TTT (in Kelvin)
|
||||
|
||||
**/<nuclide name>/reactions/reaction_<mt>/**
|
||||
|
||||
:Attributes: - **mt** (*int*) -- ENDF MT reaction number
|
||||
- **label** (*char[]*) -- Name of the reaction
|
||||
- **Q_value** (*double*) -- Q value in MeV
|
||||
- **center_of_mass** (*int*) -- Whether the reference frame for
|
||||
scattering is center-of-mass (1) or laboratory (0)
|
||||
- **n_product** (*int*) -- Number of reaction products
|
||||
|
||||
**/<nuclide name>/reactions/reaction_<mt>/<TTT>K/**
|
||||
|
||||
<TTT>K is the temperature in Kelvin, rounded to the nearest integer, of the
|
||||
temperature-dependent data set. For example, the data set corresponding to
|
||||
300 Kelvin would be located at `300K`.
|
||||
|
||||
:Datasets:
|
||||
- **xs** (*double[]*) -- Cross section values tabulated against the
|
||||
nuclide energy grid for temperature TTT (in Kelvin)
|
||||
|
||||
:Attributes:
|
||||
- **threshold_idx** (*int*) -- Index on the energy
|
||||
grid that the reaction threshold corresponds to for
|
||||
temperature TTT (in Kelvin)
|
||||
|
||||
**/<nuclide name>/reactions/reaction_<mt>/product_<j>/**
|
||||
|
||||
Reaction product data is described in :ref:`product`.
|
||||
|
||||
**/<nuclide name>/urr/<TTT>K/**
|
||||
|
||||
<TTT>K is the temperature in Kelvin, rounded to the nearest integer, of the
|
||||
temperature-dependent data set. For example, the data set corresponding to
|
||||
300 Kelvin would be located at `300K`.
|
||||
|
||||
:Attributes: - **interpolation** (*int*) -- interpolation scheme
|
||||
- **inelastic** (*int*) -- flag indicating inelastic scattering
|
||||
- **other_absorb** (*int*) -- flag indicating other absorption
|
||||
- **factors** (*int*) -- flag indicating whether tables are
|
||||
absolute or multipliers
|
||||
|
||||
:Datasets: - **energy** (*double[]*) -- Energy at which probability tables exist
|
||||
- **table** (*double[][][]*) -- Probability tables
|
||||
|
||||
**/<nuclide name>/total_nu/**
|
||||
|
||||
This special product is used to define the total number of neutrons produced
|
||||
from fission. It is formatted as a reaction product, described in
|
||||
:ref:`product`.
|
||||
|
||||
**/<nuclide name>/fission_energy_release/**
|
||||
|
||||
:Datasets: - **fragments** (:ref:`polynomial <1d_polynomial>`) -- Energy
|
||||
released in the form of fragments as a function of incident
|
||||
neutron energy.
|
||||
- **prompt_neutrons** (:ref:`polynomial <1d_polynomial>` or
|
||||
:ref:`tabulated <1d_tabulated>`) -- Energy released in the form of
|
||||
prompt neutrons as a function of incident neutron energy.
|
||||
- **delayed_neutrons** (:ref:`polynomial <1d_polynomial>`) -- Energy
|
||||
released in the form of delayed neutrons as a function of incident
|
||||
neutron energy.
|
||||
- **prompt_photons** (:ref:`polynomial <1d_polynomial>`) -- Energy
|
||||
released in the form of prompt photons as a function of incident
|
||||
neutron energy.
|
||||
- **delayed_photons** (:ref:`polynomial <1d_polynomial>`) -- Energy
|
||||
released in the form of delayed photons as a function of incident
|
||||
neutron energy.
|
||||
- **betas** (:ref:`polynomial <1d_polynomial>`) -- Energy
|
||||
released in the form of betas as a function of incident
|
||||
neutron energy.
|
||||
- **neutrinos** (:ref:`polynomial <1d_polynomial>`) -- Energy
|
||||
released in the form of neutrinos as a function of incident
|
||||
neutron energy.
|
||||
- **q_prompt** (:ref:`polynomial <1d_polynomial>` or
|
||||
:ref:`tabulated <1d_tabulated>`) -- The prompt fission Q-value
|
||||
(fragments + prompt neutrons + prompt photons - incident energy)
|
||||
- **q_recoverable** (:ref:`polynomial <1d_polynomial>` or
|
||||
:ref:`tabulated <1d_tabulated>`) -- The recoverable fission Q-value
|
||||
(Q_prompt + delayed neutrons + delayed photons + betas)
|
||||
|
||||
-------------------------------
|
||||
Thermal Neutron Scattering Data
|
||||
-------------------------------
|
||||
|
||||
**/<thermal name>/**
|
||||
|
||||
:Attributes: - **atomic_weight_ratio** (*double*) -- Mass in units of neutron masses
|
||||
- **nuclides** (*char[][]*) -- Names of nuclides for which the thermal
|
||||
scattering data applies to
|
||||
- **secondary_mode** (*char[]*) -- Indicates how the inelastic
|
||||
outgoing angle-energy distributions are represented ('equal',
|
||||
'skewed', or 'continuous').
|
||||
|
||||
**/<thermal name>/kTs/**
|
||||
|
||||
<TTT>K is the temperature in Kelvin, rounded to the nearest integer, of the
|
||||
temperature-dependent data set. For example, the data set corresponding to
|
||||
300 Kelvin would be located at `300K`.
|
||||
|
||||
:Datasets:
|
||||
- **<TTT>K** (*double*) -- kT values (in MeV) for each Temperature
|
||||
TTT (in Kelvin)
|
||||
|
||||
**/<thermal name>/elastic/<TTT>K/**
|
||||
|
||||
<TTT>K is the temperature in Kelvin, rounded to the nearest integer, of the
|
||||
temperature-dependent data set. For example, the data set corresponding to
|
||||
300 Kelvin would be located at `300K`.
|
||||
|
||||
:Datasets: - **xs** (:ref:`tabulated <1d_tabulated>`) -- Thermal inelastic
|
||||
scattering cross section for temperature TTT (in Kelvin)
|
||||
- **mu_out** (*double[][]*) -- Distribution of outgoing energies
|
||||
and angles for coherent elastic scattering for temperature TTT
|
||||
(in Kelvin)
|
||||
|
||||
**/<thermal name>/inelastic/<TTT>K/**
|
||||
|
||||
<TTT>K is the temperature in Kelvin, rounded to the nearest integer, of the
|
||||
temperature-dependent data set. For example, the data set corresponding to
|
||||
300 Kelvin would be located at `300K`.
|
||||
|
||||
:Datasets: - **xs** (:ref:`tabulated <1d_tabulated>`) -- Thermal inelastic
|
||||
scattering cross section for temperature TTT (in Kelvin)
|
||||
- **energy_out** (*double[][]*) -- Distribution of outgoing
|
||||
energies for each incoming energy for temperature TTT (in Kelvin).
|
||||
Only present if secondary mode is not continuous.
|
||||
- **mu_out** (*double[][][]*) -- Distribution of scattering cosines
|
||||
for each pair of incoming and outgoing energies. for temperature
|
||||
TTT (in Kelvin). Only present if secondary mode is not continuous.
|
||||
|
||||
If the secondary mode is continuous, the outgoing energy-angle distribution is
|
||||
given as a :ref:`correlated angle-energy distribution
|
||||
<correlated_angle_energy>`.
|
||||
|
||||
.. _product:
|
||||
|
||||
-----------------
|
||||
Reaction Products
|
||||
-----------------
|
||||
|
||||
:Object type: Group
|
||||
:Attributes: - **particle** (*char[]*) -- Type of particle
|
||||
- **emission_mode** (*char[]*) -- Emission mode (prompt, delayed,
|
||||
total)
|
||||
- **decay_rate** (*double*) -- Rate of decay in inverse seconds
|
||||
- **n_distribution** (*int*) -- Number of angle/energy
|
||||
distributions
|
||||
:Datasets:
|
||||
- **yield** (:ref:`function <1d_functions>`) -- Energy-dependent
|
||||
yield of the product.
|
||||
|
||||
:Groups:
|
||||
- **distribution_<k>** -- Formats for angle-energy distributions are
|
||||
detailed in :ref:`angle_energy`. When multiple angle-energy
|
||||
distributions occur, one dataset also may appear for each
|
||||
distribution:
|
||||
|
||||
:Datasets:
|
||||
- **applicability** (:ref:`function <1d_functions>`) --
|
||||
Probability of selecting this distribution as a function
|
||||
of incident energy
|
||||
|
||||
.. _1d_functions:
|
||||
|
||||
-------------------------
|
||||
One-dimensional Functions
|
||||
-------------------------
|
||||
|
||||
Scalar
|
||||
------
|
||||
|
||||
:Object type: Dataset
|
||||
:Datatype: *double*
|
||||
:Attributes: - **type** (*char[]*) -- 'constant'
|
||||
|
||||
.. _1d_tabulated:
|
||||
|
||||
Tabulated
|
||||
---------
|
||||
|
||||
:Object type: Dataset
|
||||
:Datatype: *double[2][]*
|
||||
:Description: x-values are listed first followed by corresponding y-values
|
||||
:Attributes: - **type** (*char[]*) -- 'Tabulated1D'
|
||||
- **breakpoints** (*int[]*) -- Region breakpoints
|
||||
- **interpolation** (*int[]*) -- Region interpolation codes
|
||||
|
||||
.. _1d_polynomial:
|
||||
|
||||
Polynomial
|
||||
----------
|
||||
|
||||
:Object type: Dataset
|
||||
:Datatype: *double[]*
|
||||
:Description: Polynomial coefficients listed in order of increasing power
|
||||
:Attributes: - **type** (*char[]*) -- 'Polynomial'
|
||||
|
||||
Coherent elastic scattering
|
||||
---------------------------
|
||||
|
||||
:Object type: Dataset
|
||||
:Datatype: *double[2][]*
|
||||
:Description: The first row lists Bragg edges and the second row lists structure
|
||||
factor cumulative sums.
|
||||
:Attributes: - **type** (*char[]*) -- 'bragg'
|
||||
|
||||
.. _angle_energy:
|
||||
|
||||
--------------------------
|
||||
Angle-Energy Distributions
|
||||
--------------------------
|
||||
|
||||
Uncorrelated Angle-Energy
|
||||
-------------------------
|
||||
|
||||
:Object type: Group
|
||||
:Attributes: - **type** (*char[]*) -- 'uncorrelated'
|
||||
:Datasets: - **angle/energy** (*double[]*) -- energies at which angle distributions exist
|
||||
- **angle/mu** (*double[3][]*) -- tabulated angular distributions for
|
||||
each energy. The first row gives :math:`\mu` values, the second row
|
||||
gives the probability density, and the third row gives the
|
||||
cumulative distribution.
|
||||
|
||||
:Attributes: - **offsets** (*int[]*) -- indices indicating where
|
||||
each angular distribution starts
|
||||
- **interpolation** (*int[]*) -- interpolation code
|
||||
for each angular distribution
|
||||
|
||||
:Groups: - **energy/** (:ref:`energy distribution <energy_distribution>`)
|
||||
|
||||
.. _correlated_angle_energy:
|
||||
|
||||
Correlated Angle-Energy
|
||||
-----------------------
|
||||
|
||||
:Object type: Group
|
||||
:Attributes: - **type** (*char[]*) -- 'correlated'
|
||||
:Datasets: - **energy** (*double[]*) -- Incoming energies at which distributions exist
|
||||
|
||||
:Attributes:
|
||||
- **interpolation** (*double[2][]*) -- Breakpoints and
|
||||
interpolation codes for incoming energy regions
|
||||
|
||||
- **energy_out** (*double[5][]*) -- Distribution of outgoing energies
|
||||
corresponding to each incoming energy. The distributions are
|
||||
flattened into a single array; the start of a given distribution
|
||||
can be determined using the ``offsets`` attribute. The first row
|
||||
gives outgoing energies, the second row gives the probability
|
||||
density, the third row gives the cumulative distribution, the
|
||||
fourth row gives interpolation codes for angular distributions, and
|
||||
the fifth row gives offsets for angular distributions.
|
||||
|
||||
:Attributes: - **offsets** (*double[]*) -- Offset for each
|
||||
distribution
|
||||
- **interpolation** (*int[]*) -- Interpolation code
|
||||
for each distribution
|
||||
- **n_discrete_lines** (*int[]*) -- Number of discrete
|
||||
lines in each distribution
|
||||
|
||||
- **mu** (*double[3][]*) -- Distribution of angular cosines
|
||||
corresponding to each pair of incoming and outgoing energies. The
|
||||
distributions are flattened into a single array; the start of a
|
||||
given distribution can be determined using offsets in the fifth row
|
||||
of the ``energy_out`` dataset. The first row gives angular cosines,
|
||||
the second row gives the probability density, and the third row
|
||||
gives the cumulative distribution.
|
||||
|
||||
Kalbach-Mann
|
||||
------------
|
||||
|
||||
:Object type: Group
|
||||
:Attributes: - **type** (*char[]*) -- 'kalbach-mann'
|
||||
:Datasets: - **energy** (*double[]*) -- Incoming energies at which distributions exist
|
||||
|
||||
:Attributes:
|
||||
- **interpolation** (*double[2][]*) -- Breakpoints and
|
||||
interpolation codes for incoming energy regions
|
||||
|
||||
- **distribution** (*double[5][]*) -- Distribution of outgoing
|
||||
energies and angles corresponding to each incoming energy. The
|
||||
distributions are flattened into a single array; the start of a
|
||||
given distribution can be determined using the ``offsets``
|
||||
attribute. The first row gives outgoing energies, the second row
|
||||
gives the probability density, the third row gives the cumulative
|
||||
distribution, the fourth row gives Kalbach-Mann precompound
|
||||
factors, and the fifth row gives Kalbach-Mann angular distribution
|
||||
slopes.
|
||||
|
||||
:Attributes: - **offsets** (*double[]*) -- Offset for each
|
||||
distribution
|
||||
- **interpolation** (*int[]*) -- Interpolation code
|
||||
for each distribution
|
||||
- **n_discrete_lines** (*int[]*) -- Number of discrete
|
||||
lines in each distribution
|
||||
|
||||
N-Body Phase Space
|
||||
------------------
|
||||
|
||||
:Object type: Group
|
||||
:Attributes: - **type** (*char[]*) -- 'nbody'
|
||||
- **total_mass** (*double*) -- Total mass of product particles
|
||||
- **n_particles** (*int*) -- Number of product particles
|
||||
- **atomic_weight_ratio** (*double*) -- Atomic weight ratio of the
|
||||
target nuclide in neutron masses
|
||||
- **q_value** (*double*) -- Q value for the reaction in MeV
|
||||
|
||||
.. _energy_distribution:
|
||||
|
||||
--------------------
|
||||
Energy Distributions
|
||||
--------------------
|
||||
|
||||
Maxwell
|
||||
-------
|
||||
|
||||
:Object type: Group
|
||||
:Attributes: - **type** (*char[]*) -- 'maxwell'
|
||||
- **u** (*double*) -- Restriction energy in MeV
|
||||
:Datasets:
|
||||
- **theta** (:ref:`tabulated <1d_tabulated>`) -- Maxwellian
|
||||
temperature as a function of energy
|
||||
|
||||
Evaporation
|
||||
-----------
|
||||
|
||||
:Object type: Group
|
||||
:Attributes: - **type** (*char[]*) -- 'evaporation'
|
||||
- **u** (*double*) -- Restriction energy in MeV
|
||||
:Datasets:
|
||||
- **theta** (:ref:`tabulated <1d_tabulated>`) -- Evaporation
|
||||
temperature as a function of energy
|
||||
|
||||
Watt Fission Spectrum
|
||||
---------------------
|
||||
|
||||
:Object type: Group
|
||||
:Attributes: - **type** (*char[]*) -- 'watt'
|
||||
- **u** (*double*) -- Restriction energy in MeV
|
||||
:Datasets: - **a** (:ref:`tabulated <1d_tabulated>`) -- Watt parameter :math:`a`
|
||||
as a function of incident energy
|
||||
- **b** (:ref:`tabulated <1d_tabulated>`) -- Watt parameter :math:`b`
|
||||
as a function of incident energy
|
||||
|
||||
Madland-Nix
|
||||
-----------
|
||||
|
||||
:Object type: Group
|
||||
:Attributes: - **type** (*char[]*) -- 'watt'
|
||||
- **efl** (*double*) -- Average energy of light fragment in eV
|
||||
- **efh** (*double*) -- Average energy of heavy fragment in eV
|
||||
|
||||
Discrete Photon
|
||||
---------------
|
||||
|
||||
:Object type: Group
|
||||
:Attributes: - **type** (*char[]*) -- 'discrete_photon'
|
||||
- **primary_flag** (*int*) -- Whether photon is a primary
|
||||
- **energy** (*double*) -- Photon energy in MeV
|
||||
- **atomic_weight_ratio** (*double*) -- Atomic weight ratio of
|
||||
target nuclide in neutron masses
|
||||
|
||||
Level Inelastic
|
||||
---------------
|
||||
|
||||
:Object type: Group
|
||||
:Attributes: - **type** (*char[]*) -- 'level'
|
||||
- **threshold** (*double*) -- Energy threshold in the laboratory
|
||||
system in MeV
|
||||
- **mass_ratio** (*double*) -- :math:`(A/(A + 1))^2`
|
||||
|
||||
Continuous Tabular
|
||||
------------------
|
||||
|
||||
:Object type: Group
|
||||
:Attributes: - **type** (*char[]*) -- 'continuous'
|
||||
:Datasets: - **energy** (*double[]*) -- Incoming energies at which distributions exist
|
||||
|
||||
:Attributes:
|
||||
- **interpolation** (*double[2][]*) -- Breakpoints and
|
||||
interpolation codes for incoming energy regions
|
||||
|
||||
- **distribution** (*double[3][]*) -- Distribution of outgoing
|
||||
energies corresponding to each incoming energy. The distributions
|
||||
are flattened into a single array; the start of a given
|
||||
distribution can be determined using the ``offsets`` attribute. The
|
||||
first row gives outgoing energies, the second row gives the
|
||||
probability density, and the third row gives the cumulative
|
||||
distribution.
|
||||
|
||||
:Attributes: - **offsets** (*double[]*) -- Offset for each
|
||||
distribution
|
||||
- **interpolation** (*int[]*) -- Interpolation code
|
||||
for each distribution
|
||||
- **n_discrete_lines** (*int[]*) -- Number of discrete
|
||||
lines in each distribution
|
||||
|
|
@ -4,7 +4,7 @@
|
|||
Summary File Format
|
||||
===================
|
||||
|
||||
The current revision of the summary file format is 1.
|
||||
The current revision of the summary file format is 4.
|
||||
|
||||
**/filetype** (*char[]*)
|
||||
|
||||
|
|
@ -129,7 +129,16 @@ The current revision of the summary file format is 1.
|
|||
|
||||
**/geometry/cells/cell <uid>/distribcell_index** (*int*)
|
||||
|
||||
Index of this cell in distribcell filter arrays.
|
||||
Index of this cell in distribcell arrays. Only present if this cell is
|
||||
listed in a distribcell filter or if it uses distributed materials.
|
||||
|
||||
**/geometry/cells/cell <uid>/paths** (*char[][]*)
|
||||
|
||||
The paths traversed through the CSG tree to reach each distribcell
|
||||
instance. This consists of the integer IDs for each universe, cell and
|
||||
lattice delimited by '->'. Each lattice cell is specified by its (x,y) or
|
||||
(x,y,z) indices. Only present if this cell is listed in a distribcell filter
|
||||
or if it uses distributed materials.
|
||||
|
||||
**/geometry/surfaces/surface <uid>/index** (*int*)
|
||||
|
||||
|
|
@ -244,90 +253,6 @@ The current revision of the summary file format is 1.
|
|||
|
||||
Names of S(:math:`\alpha`,:math:`\beta`) tables assigned to the material.
|
||||
|
||||
**/tallies/n_tallies** (*int*)
|
||||
|
||||
Number of tallies in the problem.
|
||||
|
||||
**/tallies/n_meshes** (*int*)
|
||||
|
||||
Number of meshes in the problem.
|
||||
|
||||
**/tallies/mesh <uid>/index** (*int*)
|
||||
|
||||
Index in the meshes array used internally in OpenMC.
|
||||
|
||||
**/tallies/mesh <uid>/type** (*char[]*)
|
||||
|
||||
Type of the mesh. The only valid option is currently 'regular'.
|
||||
|
||||
**/tallies/mesh <uid>/dimension** (*int[]*)
|
||||
|
||||
Number of mesh cells in each direction.
|
||||
|
||||
**/tallies/mesh <uid>/lower_left** (*double[]*)
|
||||
|
||||
Coordinates of the lower-left corner of the mesh.
|
||||
|
||||
**/tallies/mesh <uid>/upper_right** (*double[]*)
|
||||
|
||||
Coordinates of the upper-right corner of the mesh.
|
||||
|
||||
**/tallies/mesh <uid>/width** (*double[]*)
|
||||
|
||||
Width of a single mesh cell in each direction.
|
||||
|
||||
**/tallies/tally <uid>/index** (*int*)
|
||||
|
||||
Index in tallies array used internally in OpenMC.
|
||||
|
||||
**/tallies/tally <uid>/name** (*char[]*)
|
||||
|
||||
Name of the tally.
|
||||
|
||||
**/tallies/tally <uid>/n_filters** (*int*)
|
||||
|
||||
Number of filters applied to the tally.
|
||||
|
||||
**/tallies/tally <uid>/filter <j>/type** (*char[]*)
|
||||
|
||||
Type of the j-th filter. Can be 'universe', 'material', 'cell', 'cellborn',
|
||||
'surface', 'mesh', 'energy', 'energyout', or 'distribcell'.
|
||||
|
||||
**/tallies/tally <uid>/filter <j>/offset** (*int*)
|
||||
|
||||
Filter offset (used for distribcell filter).
|
||||
|
||||
**/tallies/tally <uid>/filter <j>/paths** (*char[][]*)
|
||||
|
||||
The paths traversed through the CSG tree to reach each distribcell
|
||||
instance (for 'distribcell' filters only). This consists of the integer
|
||||
IDs for each universe, cell and lattice delimited by '->'. Each lattice
|
||||
cell is specified by its (x,y) or (x,y,z) indices.
|
||||
|
||||
**/tallies/tally <uid>/filter <j>/n_bins** (*int*)
|
||||
|
||||
Number of bins for the j-th filter.
|
||||
|
||||
**/tallies/tally <uid>/filter <j>/bins** (*int[]* or *double[]*)
|
||||
|
||||
Value for each filter bin of this type.
|
||||
|
||||
**/tallies/tally <uid>/nuclides** (*char[][]*)
|
||||
|
||||
Array of nuclides to tally. Note that if no nuclide is specified in the user
|
||||
input, a single 'total' nuclide appears here.
|
||||
|
||||
**/tallies/tally <uid>/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.
|
||||
|
||||
**/tallies/tally <uid>/moment_orders** (*char[][]*)
|
||||
|
||||
Tallying moment orders for Legendre and spherical harmonic tally expansions
|
||||
(*e.g.*, 'P2', 'Y1,2', etc.).
|
||||
|
||||
**/tallies/tally <uid>/score_bins** (*char[][]*)
|
||||
|
||||
Scoring bins for the tally.
|
||||
|
|
|
|||
22
docs/source/io_formats/volume.rst
Normal file
22
docs/source/io_formats/volume.rst
Normal file
|
|
@ -0,0 +1,22 @@
|
|||
.. _io_volume:
|
||||
|
||||
==================
|
||||
Volume File Format
|
||||
==================
|
||||
|
||||
**/**
|
||||
|
||||
:Attributes: - **samples** (*int*) -- Number of samples
|
||||
- **lower_left** (*double[3]*) -- Lower-left coordinates of
|
||||
bounding box
|
||||
- **upper_right** (*double[3]*) -- Upper-right coordinates of
|
||||
bounding box
|
||||
|
||||
**/cell_<id>/**
|
||||
|
||||
:Datasets: - **volume** (*double[2]*) -- Calculated volume and its uncertainty
|
||||
in cubic centimeters
|
||||
- **nuclides** (*char[][]*) -- Names of nuclides identified in the
|
||||
cell
|
||||
- **atoms** (*double[][2]*) -- Total number of atoms of each nuclide
|
||||
and its uncertainty
|
||||
|
|
@ -8,6 +8,10 @@ This page section discusses how nonlinear diffusion acceleration (NDA) using
|
|||
coarse mesh finite difference (CMFD) is implemented into OpenMC. Before we get
|
||||
into the theory, general notation for this section is discussed.
|
||||
|
||||
Note that the methods discussed in this section are written specifically for
|
||||
continuous-energy mode but equivalent apply to the multi-group mode if the
|
||||
particle's energy is replaced with the particle's group
|
||||
|
||||
--------
|
||||
Notation
|
||||
--------
|
||||
|
|
|
|||
|
|
@ -1,16 +1,20 @@
|
|||
.. _methods_cross_sections:
|
||||
|
||||
============================
|
||||
Cross Section Representation
|
||||
============================
|
||||
=============================
|
||||
Cross Section Representations
|
||||
=============================
|
||||
|
||||
The data governing the interaction of neutrons with various nuclei are
|
||||
represented using the ACE format which is used by MCNP_ and Serpent_. ACE-format
|
||||
data can be generated with the NJOY_ nuclear data processing system which
|
||||
converts raw `ENDF/B data`_ into linearly-interpolable data as required by most
|
||||
Monte Carlo codes. The use of a standard cross section format allows for a
|
||||
direct comparison of OpenMC with other codes since the same cross section
|
||||
libraries can be used.
|
||||
----------------------
|
||||
Continuous-Energy Data
|
||||
----------------------
|
||||
|
||||
The data governing the interaction of neutrons with
|
||||
various nuclei for continous-energy problems are represented using the ACE
|
||||
format which is used by MCNP_ and Serpent_. ACE-format data can be generated
|
||||
with the NJOY_ nuclear data processing system which converts raw
|
||||
`ENDF/B data`_ into linearly-interpolable data as required by most Monte Carlo
|
||||
codes. The use of a standard cross section format allows for a direct comparison
|
||||
of OpenMC with other codes since the same cross section libraries can be used.
|
||||
|
||||
The ACE format contains continuous-energy cross sections for the following types
|
||||
of reactions: elastic scattering, fission (or first-chance fission,
|
||||
|
|
@ -24,7 +28,6 @@ accurate treatment of self-shielding in the unresolved resonance range. For
|
|||
bound scatterers, separate tables with :math:`S(\alpha,\beta,T)` scattering law
|
||||
data can be used.
|
||||
|
||||
-------------------
|
||||
Energy Grid Methods
|
||||
-------------------
|
||||
|
||||
|
|
@ -48,22 +51,23 @@ implement a method of reducing the number of energy grid searches in order to
|
|||
speed up the calculation.
|
||||
|
||||
Logarithmic Mapping
|
||||
-------------------
|
||||
+++++++++++++++++++
|
||||
|
||||
To speed up energy grid searches, OpenMC uses logarithmic mapping technique
|
||||
[Brown]_ to limit the range of energies that must be searched for each
|
||||
nuclide. The entire energy range is divided up into equal-lethargy segments, and
|
||||
the bounding energies of each segment are mapped to bounding indices on each of
|
||||
the nuclide energy grids. By default, OpenMC uses 8000 equal-lethargy segments
|
||||
as recommended by Brown.
|
||||
To speed up energy grid searches, OpenMC uses a `logarithmic mapping technique`_
|
||||
to limit the range of energies that must be searched for each nuclide. The
|
||||
entire energy range is divided up into equal-lethargy segments, and the bounding
|
||||
energies of each segment are mapped to bounding indices on each of the nuclide
|
||||
energy grids. By default, OpenMC uses 8000 equal-lethargy segments as
|
||||
recommended by Brown.
|
||||
|
||||
Other Methods
|
||||
-------------
|
||||
+++++++++++++
|
||||
|
||||
A good survey of other energy grid techniques, including unionized energy grids,
|
||||
can be found in a paper by Leppanen_.
|
||||
|
||||
---------------------------------
|
||||
.. _windowed_multipole:
|
||||
|
||||
Windowed Multipole Representation
|
||||
---------------------------------
|
||||
|
||||
|
|
@ -72,9 +76,9 @@ offers support for an experimental data format called windowed multipole (WMP).
|
|||
This data format requires less memory than pointwise cross sections, and it
|
||||
allows on-the-fly Doppler broadening to arbitrary temperature.
|
||||
|
||||
The multipole method was introduced by [Hwang]_ and the faster windowed
|
||||
multipole method by [Josey]_. In the multipole format, cross section resonances
|
||||
are represented by poles, :math:`p_j`, and residues, :math:`r_j`, in the complex
|
||||
The multipole method was introduced by Hwang_ and the faster windowed multipole
|
||||
method by Josey_. In the multipole format, cross section resonances are
|
||||
represented by poles, :math:`p_j`, and residues, :math:`r_j`, in the complex
|
||||
plane. The 0K cross sections in the resolved resonance region can be computed
|
||||
by summing up a contribution from each pole:
|
||||
|
||||
|
|
@ -137,23 +141,137 @@ scattering does not occur in the resolved resonance region. This is usually,
|
|||
but not always the case. Future library versions may eliminate this issue.
|
||||
|
||||
The data format used by OpenMC to represent windowed multipole data is specified
|
||||
in :ref:`io_data_wmp`
|
||||
in :ref:`io_data_wmp`.
|
||||
|
||||
.. only:: html
|
||||
.. _temperature_treatment:
|
||||
|
||||
.. rubric:: References
|
||||
Temperature Treatment
|
||||
---------------------
|
||||
|
||||
.. [Brown] Forrest B. Brown, "New Hash-based Energy Lookup Algorithm for Monte
|
||||
Carlo codes," LA-UR-14-24530, Los Alamos National Laboratory (2014).
|
||||
At the beginning of a simulation, OpenMC collects a list of all temperatures
|
||||
that are present in a model. It then uses this list to determine what cross
|
||||
sections to load. The data that is loaded depends on what temperature method has
|
||||
been selected. There are three methods available:
|
||||
|
||||
.. [Hwang] R. N. Hwang, "A Rigorous Pole Representation of Multilevel Cross
|
||||
Sections and Its Practical Application," *Nucl. Sci. Eng.*, **96**,
|
||||
192-209 (1987).
|
||||
:Nearest: Cross sections are loaded only if they are within a specified
|
||||
tolerance of the actual temperatures in the model.
|
||||
|
||||
.. [Josey] Colin Josey, Pablo Ducru, Benoit Forget, and Kord Smith, "Windowed
|
||||
Multipole for Cross Section Doppler Broadening," *J. Comp. Phys*,
|
||||
**307**, 715-727 (2016). http://dx.doi.org/10.1016/j.jcp.2015.08.013
|
||||
:Interpolation: Cross sections are loaded at temperatures that bound the actual
|
||||
temperatures in the model. During transport, cross sections for
|
||||
each material are calculated using statistical linear-linear
|
||||
interpolation between bounding temperature. Suppose cross
|
||||
sections are available at temperatures :math:`T_1, T_2, ...,
|
||||
T_n` and a material is assigned a temperature :math:`T` where
|
||||
:math:`T_i < T < T_{i+1}`. Statistical interpolation is applied
|
||||
as follows: a uniformly-distributed random number of the unit
|
||||
interval, :math:`\xi`, is sampled. If :math:`\xi < (T -
|
||||
T_i)/(T_{i+1} - T_i)`, then cross sections at temperature
|
||||
:math:`T_{i+1}` are used. Otherwise, cross sections at
|
||||
:math:`T_i` are used. This procedure is applied for pointwise
|
||||
cross sections in the resolved resonance range, unresolved
|
||||
resonance probability tables, and :math:`S(\alpha,\beta)`
|
||||
thermal scattering tables.
|
||||
|
||||
:Multipole: Resolved resonance cross sections are calculated on-the-fly using
|
||||
techniques/data described in :ref:`windowed_multipole`. Cross
|
||||
section data is loaded for a single temperature and is used in the
|
||||
unresolved resonance and fast energy ranges.
|
||||
|
||||
----------------
|
||||
Multi-Group Data
|
||||
----------------
|
||||
|
||||
The data governing the interaction of particles with various nuclei or materials
|
||||
are represented using a multi-group library format specific to the OpenMC code.
|
||||
The format is described in the :ref:`mgxs_lib_spec`.
|
||||
The data itself can be prepared via traditional paths or directly from a
|
||||
continuous-energy OpenMC calculation by use of the Python API as is shown in the
|
||||
:ref:`notebook_mgxs_part_iv` example notebook. This multi-group
|
||||
library consists of meta-data (such as the energy group structure) and multiple
|
||||
`xsdata` objects which contains the required microscopic or macroscopic
|
||||
multi-group data.
|
||||
|
||||
At a minimum, the library must contain the absorption cross section
|
||||
(:math:`\sigma_{a,g}`) and a scattering matrix. If the problem is an eigenvalue
|
||||
problem then all fissionable materials must also contain either
|
||||
a fission production matrix cross section
|
||||
(:math:`\nu\sigma_{f,g\rightarrow g'}`), or
|
||||
both the fission spectrum data (:math:`\chi_{g'}`) and a fission production
|
||||
cross section (:math:`\nu\sigma_{f,g}`), or, . The library must also contain
|
||||
the fission cross section (:math:`\sigma_{f,g}`) or the fission energy release
|
||||
cross section (:math:`\kappa\sigma_{f,g}`) if the associated tallies are
|
||||
required by the model using the library.
|
||||
|
||||
After a scattering collision, the outgoing particle experiences a change in both
|
||||
energy and angle. The probability of a particle resulting in a given outgoing
|
||||
energy group (`g'`) given a certain incoming energy group (`g`) is provided
|
||||
by the scattering matrix data. The angular information can be expressed either
|
||||
via Legendre expansion of the particle's change-in-angle (:math:`\mu`), a
|
||||
tabular representation of the probability distribution function of :math:`\mu`,
|
||||
or a histogram representation of the same PDF. The formats used to
|
||||
represent these are described in the :ref:`mgxs_lib_spec`.
|
||||
|
||||
Unlike the continuous-energy mode, the multi-group mode does not explicitly
|
||||
track particles produced from scattering multiplication (i.e., :math:`(n,xn)`)
|
||||
reactions. These are instead accounted for by adjusting the weight of the
|
||||
particle after the collision such that the correct total weight is maintained.
|
||||
The weight adjustment factor is optionally provided by the `multiplicity` data
|
||||
which is required to be provided in the form of a group-wise matrix.
|
||||
This data is provided as a group-wise matrix since the probability of producing
|
||||
multiple particles in a scattering reaction depends on both the incoming energy,
|
||||
`g`, and the sampled outgoing energy, `g'`. This data represents the average
|
||||
number of particles emitted from a scattering reaction, given a scattering
|
||||
reaction has occurred:
|
||||
|
||||
.. math::
|
||||
|
||||
multiplicity_{g \rightarrow g'} = \frac{\nu_{scatter}\sigma_{s,g \rightarrow g'}}{
|
||||
\sigma_{s,g \rightarrow g'}}
|
||||
|
||||
If this scattering multiplication information is not provided in the library
|
||||
then no weight adjustment will be performed. This is equivalent to neglecting
|
||||
any additional particles produced in scattering multiplication reactions.
|
||||
However, this assumption will result in a loss of accuracy since the total
|
||||
particle population would not be conserved. This reduction in accuracy due to
|
||||
the loss in particle conservation can be mitigated by reducing the absorption
|
||||
cross section as needed to maintain particle conservation. This adjustment can
|
||||
be done when generating the library, or by OpenMC. To have OpenMC perform the
|
||||
adjustment, the total cross section (:math:`\sigma_{t,g}`) must be provided.
|
||||
With this information, OpenMC will then adjust the absorption cross section as
|
||||
follows:
|
||||
|
||||
.. math::
|
||||
|
||||
\sigma_{a,g} = \sigma_{t,g} - \sum_{g'}\nu_{scatter}\sigma_{s,g \rightarrow g'}
|
||||
|
||||
The above method is the same as is usually done with most deterministic solvers.
|
||||
Note that this method is less accurate than using the scattering multiplication
|
||||
weight adjustment since simply reducing the absorption cross section does not
|
||||
include any information about the outgoing energy of the particles produced in
|
||||
these reactions.
|
||||
|
||||
All of the data discussed in this section can be provided to the code
|
||||
independent of the particle's direction of motion (i.e., isotropic), or the data
|
||||
can be provided as a tabular distribution of the polar and azimuthal particle
|
||||
direction angles. The isotropic representation is the most commonly used,
|
||||
however inaccuracies are to be expected especially near material interfaces
|
||||
where a material has a very large cross sections relative to the other material
|
||||
(as can be expected in the resonance range). The angular representation can be
|
||||
used to minimize this error.
|
||||
|
||||
Finally, the above options for representing the physics do not have to be
|
||||
consistent across the problem. The number of groups and the structure, however,
|
||||
does have to be consistent across the data sets. That is to say that each
|
||||
microscopic or macroscopic data set does not have to apply the same scattering
|
||||
expansion, treatment of multiplicity or angular representation of the cross
|
||||
sections. This allows flexibility for the model to use highly anisotropic
|
||||
scattering information in the water while the fuel can be simulated with linear
|
||||
or even isotropic scattering.
|
||||
|
||||
.. _logarithmic mapping technique:
|
||||
https://laws.lanl.gov/vhosts/mcnp.lanl.gov/pdf_files/la-ur-14-24530.pdf
|
||||
.. _Hwang: http://www.ans.org/pubs/journals/nse/a_16381
|
||||
.. _Josey: http://dx.doi.org/10.1016/j.jcp.2015.08.013
|
||||
.. _MCNP: http://mcnp.lanl.gov
|
||||
.. _Serpent: http://montecarlo.vtt.fi
|
||||
.. _NJOY: http://t2.lanl.gov/codes.shtml
|
||||
|
|
|
|||
|
|
@ -205,7 +205,7 @@ traveling in its current direction, it will not hit the surface. The complete
|
|||
derivation for different types of surfaces used in OpenMC will be presented in
|
||||
the following sections.
|
||||
|
||||
Since :math:f(x,y,z)` in general is quadratic in :math:`x`, :math:`y`, and
|
||||
Since :math:`f(x,y,z)` in general is quadratic in :math:`x`, :math:`y`, and
|
||||
:math:`z`, this implies that :math:`f(x_0 + du_0, y + dv_0, z + dw_0)` is
|
||||
quadratic in :math:`d`. Thus we expect at most two real solutions to
|
||||
:eq:`dist-to-boundary-1`. If no solutions to :eq:`dist-to-boundary-1` exist or
|
||||
|
|
@ -265,6 +265,8 @@ Again, we need to check whether the denominator is zero. If so, this means that
|
|||
the particle's direction of flight is parallel to the plane and it will
|
||||
therefore never hit the plane.
|
||||
|
||||
.. _cylinder_distance:
|
||||
|
||||
Cylinder Parallel to an Axis
|
||||
----------------------------
|
||||
|
||||
|
|
@ -366,7 +368,74 @@ will then be either both positive or both negative. If they are both positive,
|
|||
the smaller (closer) one will be the solution with a negative sign on the square
|
||||
root of the discriminant.
|
||||
|
||||
.. TODO: Need to add derivation for x-cone, y-cone, and z-cone.
|
||||
Cone Parallel to an Axis
|
||||
------------------------
|
||||
|
||||
The equation for a cone parallel to, for example, the x-axis is :math:`(y -
|
||||
y_0)^2 + (z - z_0)^2 = R^2(x - x_0)^2`. Thus, we need to solve :math:`(y + dv -
|
||||
y_0)^2 + (z + dw - z_0)^2 = R^2(x + du - x_0)^2`. Let us define :math:`\bar{x} =
|
||||
x - x_0`, :math:`\bar{y} = y - y_0`, and :math:`\bar{z} = z - z_0`. We then have
|
||||
|
||||
.. math::
|
||||
:label: dist-xcone-1
|
||||
|
||||
(\bar{y} + dv)^2 + (\bar{z} + dw)^2 = R^2(\bar{x} + du)^2
|
||||
|
||||
Expanding equation :eq:`dist-xcone-1` and rearranging terms, we obtain
|
||||
|
||||
.. math::
|
||||
:label: dist-xcylinder-2
|
||||
|
||||
(v^2 + w^2 - R^2u^2) d^2 + 2 (\bar{y}v + \bar{z}w - R^2\bar{x}u) d +
|
||||
(\bar{y}^2 + \bar{z}^2 - R^2\bar{x}^2) = 0
|
||||
|
||||
Defining the terms
|
||||
|
||||
.. math::
|
||||
:label: dist-quadric-terms
|
||||
|
||||
a = v^2 + w^2 - R^2u^2
|
||||
|
||||
k = \bar{y}v + \bar{z}w - R^2\bar{x}u
|
||||
|
||||
c = \bar{y}^2 + \bar{z}^2 - R^2\bar{x}^2
|
||||
|
||||
we then have the simple quadratic equation :math:`ad^2 + 2kd + c = 0` which can
|
||||
be solved as described in :ref:`cylinder_distance`.
|
||||
|
||||
General Quadric
|
||||
---------------
|
||||
|
||||
The equation for a general quadric surface is :math:`Ax^2 + By^2 + Cz^2 + Dxy +
|
||||
Eyz + Fxz + Gx + Hy + Jz + K = 0`. Thus, we need to solve the equation
|
||||
|
||||
.. math::
|
||||
:label: dist-quadric-1
|
||||
|
||||
A(x+du)^2 + B(y+dv)^2 + C(z+dw)^2 + D(x+du)(y+dv) + E(y+dv)(z+dw) + \\
|
||||
F(x+du)(z+dw) + G(x+du) + H(y+dv) + J(z+dw) + K = 0
|
||||
|
||||
Expanding equation :eq:`dist-quadric-1` and rearranging terms, we obtain
|
||||
|
||||
.. math::
|
||||
:label: dist-quadric-2
|
||||
|
||||
d^2(uv + vw + uw) + 2d(Aux + Bvy + Cwx + (D(uv + vx) + E(vz + wy) + \\
|
||||
F(wx + uz))/2) + (x(Ax + Dy) + y(By + Ez) + z(Cz + Fx)) = 0
|
||||
|
||||
Defining the terms
|
||||
|
||||
.. math::
|
||||
:label: dist-quadric-terms
|
||||
|
||||
a = uv + vw + uw
|
||||
|
||||
k = Aux + Bvy + Cwx + (D(uv + vx) + E(vz + wy) + F(wx + uz))/2
|
||||
|
||||
c = x(Ax + Dy) + y(By + Ez) + z(Cz + Fx)
|
||||
|
||||
we then have the simple quadratic equation :math:`ad^2 + 2kd + c = 0` which can
|
||||
be solved as described in :ref:`cylinder_distance`.
|
||||
|
||||
.. _find-cell:
|
||||
|
||||
|
|
@ -437,6 +506,8 @@ where :math:`(x_0, y_0, z_0)` are the coordinates to the lower-left-bottom
|
|||
corner of the lattice, and :math:`p_0, p_1, p_2` are the pitches along the
|
||||
:math:`x`, :math:`y`, and :math:`z` axes, respectively.
|
||||
|
||||
.. _hexagonal_indexing:
|
||||
|
||||
Hexagonal Lattice Indexing
|
||||
--------------------------
|
||||
|
||||
|
|
@ -808,6 +879,18 @@ form of the solution:
|
|||
|
||||
w' = w + \frac{2 (\bar{x}u + \bar{y}v - R^2\bar{z}w)}{R^2 (1 + R^2) \bar{z}}
|
||||
|
||||
General Quadric
|
||||
---------------
|
||||
|
||||
A general quadric surface has the form :math:`f(x,y,z) = Ax^2 + By^2 + Cz^2 +
|
||||
Dxy + Eyz + Fxz + Gx + Hy + Jz + K = 0`. Thus, the gradient to the surface is
|
||||
|
||||
.. math::
|
||||
:label: reflection-quadric-grad
|
||||
|
||||
\nabla f = \left ( \begin{array}{c} 2Ax + Dy + Fz + G \\ 2By + Dx + Ez + H
|
||||
\\ 2Cz + Ey + Fx + J \end{array} \right ).
|
||||
|
||||
|
||||
.. _constructive solid geometry: http://en.wikipedia.org/wiki/Constructive_solid_geometry
|
||||
.. _surfaces: http://en.wikipedia.org/wiki/Surface
|
||||
|
|
|
|||
|
|
@ -8,7 +8,7 @@ The physical process by which a population of particles evolves over time is
|
|||
governed by a number of `probability distributions`_. For instance, given a
|
||||
particle traveling through some material, there is a probability distribution
|
||||
for the distance it will travel until its next collision (an exponential
|
||||
distribution). Then, when it collides with a nucleus, there is associated
|
||||
distribution). Then, when it collides with a nucleus, there is an associated
|
||||
probability of undergoing each possible reaction with that nucleus. While the
|
||||
behavior of any single particle is unpredictable, the average behavior of a
|
||||
large population of particles originating from the same source is well defined.
|
||||
|
|
@ -45,15 +45,20 @@ following steps:
|
|||
|
||||
- Initialize the pseudorandom number generator.
|
||||
|
||||
- Read ACE format cross sections specified in the problem.
|
||||
- Read the contiuous-energy or multi-group cross section data specified in
|
||||
the problem.
|
||||
|
||||
- If using a special energy grid treatment such as a union energy grid or
|
||||
lethargy bins, that must be initialized as well.
|
||||
lethargy bins, that must be initialized as well in a continuous-energy
|
||||
problem.
|
||||
|
||||
- In a multi-group problem, individual nuclide cross section information is
|
||||
combined to produce material-specific cross section data.
|
||||
|
||||
- In a fixed source problem, source sites are sampled from the specified
|
||||
source. In an eigenvalue problem, source sites are sampled from some initial
|
||||
source distribution or from a source file. The source sites consist of
|
||||
coordinates, a direction, and an energy.
|
||||
source. In an eigenvalue problem, source sites are sampled from some
|
||||
initial source distribution or from a source file. The source sites
|
||||
consist of coordinates, a direction, and an energy.
|
||||
|
||||
Once initialization is complete, the actual transport simulation can
|
||||
proceed. The life of a single particle will proceed as follows:
|
||||
|
|
@ -95,6 +100,10 @@ proceed. The life of a single particle will proceed as follows:
|
|||
|
||||
P(i) = \frac{\Sigma_{t,i}}{\Sigma_t}.
|
||||
|
||||
Note that the above selection of collided nuclide only applies to
|
||||
continuous-energy simulations as multi-group simulations use nuclide
|
||||
data which has already been combined in to material-specific data.
|
||||
|
||||
8. Once the specific nuclide is sampled, the random samples a reaction for
|
||||
that nuclide based on the microscopic cross sections. If the microscopic
|
||||
cross section for some reaction :math:`x` is :math:`\sigma_x` and the total
|
||||
|
|
@ -105,13 +114,20 @@ proceed. The life of a single particle will proceed as follows:
|
|||
|
||||
P(x) = \frac{\sigma_x}{\sigma_t}.
|
||||
|
||||
Since multi-group simulations use material-specific data, the above is
|
||||
performed with those material multi-group cross sections (i.e.,
|
||||
macroscopic cross sections for the material) instead of microscopic
|
||||
cross sections for the nuclide).
|
||||
|
||||
9. If the sampled reaction is elastic or inelastic scattering, the outgoing
|
||||
energy and angle is sampled from the appropriate distribution. Reactions
|
||||
of type :math:`(n,xn)` are treated as scattering and the weight of the
|
||||
particle is increased by the multiplicity of the reaction. The particle
|
||||
then continues from step 3. If the reaction is absorption or fission, the
|
||||
particle dies and if necessary, fission sites are created and stored in the
|
||||
fission bank.
|
||||
energy and angle is sampled from the appropriate distribution. In
|
||||
continuous-energy simulation, reactions of type :math:`(n,xn)` are treated
|
||||
as scattering and any additional particles which may be created are added
|
||||
to a secondary particle bank to be tracked later. In a multi-group
|
||||
simulation, this secondary bank is not used but the particle weight is
|
||||
increased accordingly. The original particle then continues from step 3.
|
||||
If the reaction is absorption or fission, the particle dies and if
|
||||
necessary, fission sites are created and stored in the fission bank.
|
||||
|
||||
After all particles have been simulated, there are a few final tasks that must
|
||||
be performed before the run is finished. This include the following:
|
||||
|
|
|
|||
|
|
@ -4,6 +4,12 @@
|
|||
Physics
|
||||
=======
|
||||
|
||||
There are limited differences between physics treatments used in the
|
||||
continuous-energy and multi-group modes. If distinctions are necessary, each
|
||||
of the following sections will provide an explanation of the differences.
|
||||
Otherwise, replacing any references of the particle's energy (`E`) with
|
||||
references to the particle's energy group (`g`) will suffice.
|
||||
|
||||
-----------------------------------
|
||||
Sampling Distance to Next Collision
|
||||
-----------------------------------
|
||||
|
|
@ -79,6 +85,10 @@ originating from :math:`(n,\gamma)` and other reactions.
|
|||
Elastic Scattering
|
||||
------------------
|
||||
|
||||
Note that the multi-group mode makes no distinction between elastic or
|
||||
inelastic scattering reactions. The spceific multi-group scattering
|
||||
implementation is discussed in the :ref:`multi-group-scatter` section.
|
||||
|
||||
Elastic scattering refers to the process by which a neutron scatters off a
|
||||
nucleus and does not leave it in an excited. It is referred to as "elastic"
|
||||
because in the center-of-mass system, the neutron does not actually lose
|
||||
|
|
@ -170,6 +180,10 @@ final direction in the lab system.
|
|||
Inelastic Scattering
|
||||
--------------------
|
||||
|
||||
Note that the multi-group mode makes no distinction between elastic or
|
||||
inelastic scattering reactions. The spceific multi-group scattering
|
||||
implementation is discussed in the :ref:`multi-group-scatter` section.
|
||||
|
||||
The major algorithms for inelastic scattering were described in previous
|
||||
sections. First, a scattering cosine is sampled using the algorithms in
|
||||
:ref:`sample-angle`. Then an outgoing energy is sampled using the algorithms in
|
||||
|
|
@ -186,12 +200,69 @@ secondary photons from nuclear de-excitation are tracked in OpenMC.
|
|||
:math:`(n,xn)` Reactions
|
||||
------------------------
|
||||
|
||||
Note that the multi-group mode makes no distinction between elastic or
|
||||
inelastic scattering reactions. The specific multi-group scattering
|
||||
implementation is discussed in the :ref:`multi-group-scatter` section.
|
||||
|
||||
These types of reactions are just treated as inelastic scattering and as such
|
||||
are subject to the same procedure as described in :ref:`inelastic-scatter`. For
|
||||
reactions with integral multiplicity, e.g., :math:`(n,2n)`, an appropriate
|
||||
number of secondary neutrons are created. For reactions that have a multiplicity
|
||||
given as a function of the incoming neutron energy (which occasionally occurs
|
||||
for MT=5), the weight of the outgoing neutron is multiplied by the multiplcity.
|
||||
for MT=5), the weight of the outgoing neutron is multiplied by the multiplicity.
|
||||
|
||||
.. _multi-group-scatter:
|
||||
|
||||
----------------------
|
||||
Multi-Group Scattering
|
||||
----------------------
|
||||
|
||||
In multi-group mode, a scattering collision requires that the outgoing energy
|
||||
group of the simulated particle be selected from a probability distribution,
|
||||
the change-in-angle selected from a probability distribution according to
|
||||
the outgoing energy group, and finally the particle's weight adjusted again
|
||||
according to the outgoing energy group.
|
||||
|
||||
The first step in selecting an outgoing energy group for a particle in a given
|
||||
incoming energy group is to select a random number (:math:`\xi`) between 0 and
|
||||
1. This number is then compared to the cumulative distribution function
|
||||
produced from the outgoing group (`g'`) data for the given incoming group (`g`):
|
||||
|
||||
.. math::
|
||||
CDF = \sum_{g'=0}^{h}\Sigma_{s,g \rightarrow g'}
|
||||
|
||||
If the scattering data is represented as a Legendre expansion, then the
|
||||
value of :math:`\Sigma_{s,g \rightarrow g'}` above is the 0th order forthe
|
||||
given group transfer. If the data is provided as tabular or histogram data, then
|
||||
:math:`\Sigma_{s,g \rightarrow g'}` is the sum of all bins of data for a given
|
||||
`g` and `g'` pair.
|
||||
|
||||
Now that the outgoing energy is known the change-in-angle, :math:`\mu` can be
|
||||
determined. If the data is provided as a Legendre expansion, this is done by
|
||||
rejection sampling of the probability distribution represented by the Legendre
|
||||
series. For efficiency, the selected values of the PDF (:math:`f(\mu)`) are
|
||||
chosen to be between 0 and the maximum value of :math:`f(\mu)` in the domain of
|
||||
-1 to 1. Note that this sampling scheme automatically forces negative values of
|
||||
the :math:`f(\mu)` probability distribution function to be treated as zero
|
||||
probabilities.
|
||||
|
||||
If the angular data is instead provided as a tabular representation, then the
|
||||
value of :math:`\mu` is selected as described in the :ref:`angle-tabular`
|
||||
section with a linear-linear interpolation scheme.
|
||||
|
||||
If the angular data is provided as a histogram representation, then
|
||||
the value of :math:`\mu` is selected in a similar fashion to that described for
|
||||
the selection of the outgoing energy (since the energy group representation is
|
||||
simply a histogram representation) except the CDF is composed of the angular
|
||||
bins and not the energy groups. However, since we are interested in a specific
|
||||
value of :math:`\mu` instead of a group, then an angle selected from a uniform
|
||||
distribution within from the chosen angular bin.
|
||||
|
||||
The final step in the scattering treatment is to adjust the weight of the
|
||||
neutron to account for any production of neutrons due to :math:`(n,xn)`
|
||||
reactions. This data is obtained from the multiplicity data provided in the
|
||||
multi-group cross section library for the material of interest.
|
||||
The scaled value will default to 1.0 if no value is provided in the library.
|
||||
|
||||
.. _fission:
|
||||
|
||||
|
|
@ -208,9 +279,9 @@ idiosyncrasies in treating fission. In an eigenvalue calculation, secondary
|
|||
neutrons from fission are only "banked" for use in the next generation rather
|
||||
than being tracked as secondary neutrons from elastic and inelastic scattering
|
||||
would be. On top of this, fission is sometimes broken into first-chance fission,
|
||||
second-chance fission, etc. An ACE table either lists the partial fission
|
||||
reactions with secondary energy distributions for each one, or a total fission
|
||||
reaction with a single secondary energy distribution.
|
||||
second-chance fission, etc. The nuclear data file either lists the partial
|
||||
fission reactions with secondary energy distributions for each one, or a total
|
||||
fission reaction with a single secondary energy distribution.
|
||||
|
||||
When a fission reaction is sampled in OpenMC (either total fission or, if data
|
||||
exists, first- or second-chance fission), the following algorithm is used to
|
||||
|
|
@ -219,7 +290,7 @@ number of prompt and delayed neutrons must be determined to decide whether the
|
|||
secondary neutrons will be prompt or delayed. This is important because delayed
|
||||
neutrons have a markedly different spectrum from prompt neutrons, one that has a
|
||||
lower average energy of emission. The total number of neutrons emitted
|
||||
:math:`\nu_t` is given as a function of incident energy in the ACE format. Two
|
||||
:math:`\nu_t` is given as a function of incident energy in the ENDF format. Two
|
||||
representations exist for :math:`\nu_t`. The first is a polynomial of order
|
||||
:math:`N` with coefficients :math:`c_0,c_1,\dots,c_N`. If :math:`\nu_t` has this
|
||||
format, we can evaluate it at incoming energy :math:`E` by using the equation
|
||||
|
|
@ -271,22 +342,57 @@ position of the collision site are stored in an array called the fission
|
|||
bank. In a subsequent generation, these fission bank sites are used as starting
|
||||
source sites.
|
||||
|
||||
-----------------------------------------
|
||||
Secondary Angles and Energy Distributions
|
||||
-----------------------------------------
|
||||
The above description is similar for the multi-group mode except the data are
|
||||
provided as group-wise data instead of in a continuous-energy format. In this
|
||||
case, the outgoing energy of the fission neutrons are represented as histograms
|
||||
by way of either the nu-fission matrix or chi vector.
|
||||
|
||||
For any reactions with secondary neutrons, it is necessary to sample secondary
|
||||
angle and energy distributions. This includes elastic and inelastic scattering,
|
||||
fission, and :math:`(n,xn)` reactions. In some cases, the angle and energy
|
||||
distributions may be specified separately, and in other cases, they may be
|
||||
specified as a correlated angle-energy distribution. In the following sections,
|
||||
we will outline the methods used to sample secondary distributions as well as
|
||||
how they are used to modify the state of a particle.
|
||||
------------------------------------
|
||||
Secondary Angle-Energy Distributions
|
||||
------------------------------------
|
||||
|
||||
Note that this section is specific to continuous-energy mode since the
|
||||
multi-group scattering process has already been described including the
|
||||
secondary energy and angle sampling.
|
||||
|
||||
For a reaction with secondary products, it is necessary to determine the
|
||||
outgoing angle and energy of the products. For any reaction other than elastic
|
||||
and level inelastic scattering, the outgoing energy must be determined based on
|
||||
tabulated or parameterized data. The `ENDF-6 Format`_ specifies a variety of
|
||||
ways that the secondary energy distribution can be represented. ENDF File 5
|
||||
contains uncorrelated energy distribution whereas ENDF File 6 contains
|
||||
correlated energy-angle distributions. The ACE format specifies its own
|
||||
representations based loosely on the formats given in ENDF-6. OpenMC's HDF5
|
||||
nuclear data files use a combination of ENDF and ACE distributions; in this
|
||||
section, we will describe how the outgoing angle and energy of secondary
|
||||
particles are sampled.
|
||||
|
||||
One of the subtleties in the nuclear data format is the fact that a single
|
||||
reaction product can have multiple angle-energy distributions. This is mainly
|
||||
useful for reactions with multiple products of the same type in the exit channel
|
||||
such as :math:`(n,2n)` or :math:`(n,3n)`. In these types of reactions, each
|
||||
neutron is emitted corresponding to a different excitation level of the compound
|
||||
nucleus, and thus in general the neutrons will originate from different energy
|
||||
distributions. If multiple angle-energy distributions are present, they are
|
||||
assigned incoming-energy-dependent probabilities that can then be used to
|
||||
randomly select one.
|
||||
|
||||
Once a distribution has been selected, the procedure for determining the
|
||||
outgoing angle and energy will depend on the type of the distribution.
|
||||
|
||||
Uncorrelated Angle-Energy Distributions
|
||||
---------------------------------------
|
||||
|
||||
The first set of distributions we will look at are uncorrelated angle-energy
|
||||
distributions, where angle and energy are specified separately. For these
|
||||
distributions, OpenMC first samples the angular distribution as described
|
||||
:ref:`sample-angle` and then samples an energy as described in
|
||||
:ref:`sample-energy`.
|
||||
|
||||
.. _sample-angle:
|
||||
|
||||
Sampling Secondary Angle Distributions
|
||||
--------------------------------------
|
||||
Sampling Angular Distributions
|
||||
++++++++++++++++++++++++++++++
|
||||
|
||||
For elastic scattering, it is only necessary to specific a secondary angle
|
||||
distribution since the outgoing energy can be determined analytically. Other
|
||||
|
|
@ -294,15 +400,14 @@ reactions may also have separate secondary angle and secondary energy
|
|||
distributions that are uncorrelated. In these cases, the secondary angle
|
||||
distribution is represented as either
|
||||
|
||||
- An Isotropic angular distribution,
|
||||
- An equiprobable distribution with 32 bins, or
|
||||
- An isotropic angular distribution,
|
||||
- A tabular distribution.
|
||||
|
||||
Isotropic Angular Distribution
|
||||
++++++++++++++++++++++++++++++
|
||||
^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^
|
||||
|
||||
In the first case, no data needs to be stored on the ACE table, and the cosine
|
||||
of the scattering angle is simply calculated as
|
||||
In the first case, no data is stored in the nuclear data file, and the cosine of
|
||||
the scattering angle is simply calculated as
|
||||
|
||||
.. math::
|
||||
:label: isotropic-angle
|
||||
|
|
@ -312,42 +417,17 @@ of the scattering angle is simply calculated as
|
|||
where :math:`\mu` is the cosine of the scattering angle and :math:`\xi` is a
|
||||
random number sampled uniformly on :math:`[0,1)`.
|
||||
|
||||
Equiprobable Angle Bin Distribution
|
||||
+++++++++++++++++++++++++++++++++++
|
||||
|
||||
For a 32 equiprobable bin distribution, we select a random number :math:`\xi` to
|
||||
sample a cosine bin :math:`i` such that
|
||||
|
||||
.. math::
|
||||
:label: equiprobable-bin
|
||||
|
||||
i = 1 + \lfloor 32\xi \rfloor.
|
||||
|
||||
The same random number can then also be used to interpolate between neighboring
|
||||
:math:`\mu` values to get the final scattering cosine:
|
||||
|
||||
.. math::
|
||||
:label: equiprobable-cosine
|
||||
|
||||
\mu = \mu_i + (32\xi - i) (\mu_{i+1} - \mu_i)
|
||||
|
||||
where :math:`\mu_i` is the :math:`i`-th scattering cosine.
|
||||
|
||||
.. _angle-tabular:
|
||||
|
||||
Tabular Angular Distribution
|
||||
++++++++++++++++++++++++++++
|
||||
^^^^^^^^^^^^^^^^^^^^^^^^^^^^
|
||||
|
||||
As the `MCNP Manual`_ points out, using an equiprobable bin distribution works
|
||||
well for high-probability regions of the scattering cosine probability, but for
|
||||
low-probability regions it is not very accurate. Thus, a more accurate method is
|
||||
to represent the scattering cosine with a tabular distribution. In this case, we
|
||||
have a table of cosines and their corresponding values for a probability
|
||||
distribution function and cumulative distribution function. For each incoming
|
||||
neutron energy :math:`E_i`, let us call :math:`p_{i,j}` the j-th value in the
|
||||
probability distribution function and :math:`c_{i,j}` the j-th value in the
|
||||
cumulative distribution function. We first find the interpolation factor on the
|
||||
incoming energy grid:
|
||||
In this case, we have a table of cosines and their corresponding values for a
|
||||
probability distribution function and cumulative distribution function. For each
|
||||
incoming neutron energy :math:`E_i`, let us call :math:`p_{i,j}` the j-th value
|
||||
in the probability distribution function and :math:`c_{i,j}` the j-th value in
|
||||
the cumulative distribution function. We first find the interpolation factor on
|
||||
the incoming energy grid:
|
||||
|
||||
.. math::
|
||||
:label: interpolation-factor
|
||||
|
|
@ -465,89 +545,11 @@ linear-linear interpolation:
|
|||
|
||||
.. _sample-energy:
|
||||
|
||||
Sampling Secondary Energy and Correlated Angle/Energy Distributions
|
||||
-------------------------------------------------------------------
|
||||
Sampling Energy Distributions
|
||||
+++++++++++++++++++++++++++++
|
||||
|
||||
For a reaction with secondary neutrons, it is necessary to determine the
|
||||
outgoing energy of the neutrons. For any reaction other than elastic scattering,
|
||||
the outgoing energy must be determined based on tabulated or parameterized
|
||||
data. The `ENDF-6 Format`_ specifies a variety of ways that the secondary energy
|
||||
distribution can be represented. ENDF File 5 contains uncorrelated energy
|
||||
distribution where ENDF File 6 contains correlated energy-angle
|
||||
distributions. The ACE format specifies its own representations based loosely on
|
||||
the formats given in ENDF-6. In this section, we will describe how the outgoing
|
||||
energy of secondary particles is determined based on each ACE law.
|
||||
|
||||
One of the subtleties in the ACE format is the fact that a single reaction can
|
||||
have multiple secondary energy distributions. This is mainly useful for
|
||||
reactions with multiple neutrons in the exit channel such as :math:`(n,2n)` or
|
||||
:math:`(n,3n)`. In these types of reactions, each neutron is emitted
|
||||
corresponding to a different excitation level of the compound nucleus, and thus
|
||||
in general the neutrons will originate from different energy distributions. If
|
||||
multiple energy distributions are present, they are assigned probabilities that
|
||||
can then be used to randomly select one.
|
||||
|
||||
Once a secondary energy distribution has been sampled, the procedure for
|
||||
determining the outgoing energy will depend on which ACE law has been specified
|
||||
for the data.
|
||||
|
||||
.. _ace-law-1:
|
||||
|
||||
ACE Law 1 - Tabular Equiprobable Energy Bins
|
||||
++++++++++++++++++++++++++++++++++++++++++++
|
||||
|
||||
In the tabular equiprobable bin representation, an array of equiprobable
|
||||
outgoing energy bins is given for a number of incident energies. While the
|
||||
representation itself is simple, the complexity lies in how one interpolates
|
||||
between incident as well as outgoing energies on such a table. If one performs
|
||||
simple interpolation between tables for neighboring incident energies, it is
|
||||
possible that the resulting energies would violate laws governing the
|
||||
kinematics, i.e. the outgoing energy may be outside the range of available
|
||||
energy in the reaction.
|
||||
|
||||
To avoid this situation, the accepted practice is to use a process known as
|
||||
scaled interpolation [Doyas]_. First, we find the tabulated incident energies
|
||||
which bound the actual incoming energy of the particle, i.e. find :math:`i` such
|
||||
that :math:`E_i < E < E_{i+1}` and calculate the interpolation factor :math:`f`
|
||||
via :eq:`interpolation-factor`. Then, we interpolate between the minimum and
|
||||
maximum energies of the outgoing energy distributions corresponding to
|
||||
:math:`E_i` and :math:`E_{i+1}`:
|
||||
|
||||
.. math::
|
||||
:label: ace-law-1-minmax
|
||||
|
||||
E_{min} = E_{i,1} + f ( E_{i+1,1} - E_i ) \\
|
||||
E_{max} = E_{i,M} + f ( E_{i+1,M} - E_M )
|
||||
|
||||
where :math:`E_{min}` and :math:`E_{max}` are the minimum and maximum outgoing
|
||||
energies of a scaled distribution, :math:`E_{i,j}` is the j-th outgoing energy
|
||||
corresponding to the incoming energy :math:`E_i`, and :math:`M` is the number of
|
||||
outgoing energy bins. Next, statistical interpolation is performed to choose
|
||||
between using the outgoing energy distributions corresponding to energy
|
||||
:math:`E_i` and :math:`E_{i+1}`. Let :math:`\ell` be the chosen table where
|
||||
:math:`\ell = i` if :math:`\xi_1 > f` and :math:`\ell = i + 1` otherwise, and
|
||||
:math:`\xi_1` is a random number. Now, we randomly sample an equiprobable
|
||||
outgoing energy bin :math:`j` and interpolate between successive values on the
|
||||
outgoing energy distribution:
|
||||
|
||||
.. math::
|
||||
:label: ace-law-1-intermediate
|
||||
|
||||
\hat{E} = E_{\ell,j} + \xi_2 (E_{\ell,j+1} - E_{\ell,j})
|
||||
|
||||
where :math:`\xi_2` is a random number sampled uniformly on :math:`[0,1)`. Since
|
||||
this outgoing energy may violate reaction kinematics, we then scale it to the
|
||||
minimum and maximum energies we calculated earlier to get the final outgoing
|
||||
energy:
|
||||
|
||||
.. math::
|
||||
:label: ace-law-1-energy
|
||||
|
||||
E' = E_{min} + \frac{\hat{E} - E_{\ell,1}}{E_{\ell,M} - E_{\ell,1}}
|
||||
(E_{max} - E_{min})
|
||||
|
||||
ACE Law 3 - Inelastic Level Scattering
|
||||
++++++++++++++++++++++++++++++++++++++
|
||||
Inelastic Level Scattering
|
||||
^^^^^^^^^^^^^^^^^^^^^^^^^^
|
||||
|
||||
It can be shown (see Foderaro_) that in inelastic level scattering, the outgoing
|
||||
energy of the neutron :math:`E'` can be related to the Q-value of the reaction
|
||||
|
|
@ -560,31 +562,50 @@ and the incoming energy:
|
|||
|
||||
where :math:`A` is the mass of the target nucleus measured in neutron masses.
|
||||
|
||||
.. _ace-law-4:
|
||||
.. _continuous-tabular:
|
||||
|
||||
ACE Law 4 - Continuous Tabular Distribution
|
||||
+++++++++++++++++++++++++++++++++++++++++++
|
||||
Continuous Tabular Distribution
|
||||
^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^
|
||||
|
||||
This representation is very similar to :ref:`ace-law-1` except that instead of
|
||||
equiprobable outgoing energy bins, the outgoing energy distribution for each
|
||||
incoming energy is represented with a probability distribution function. For
|
||||
each incoming neutron energy :math:`E_i`, let us call :math:`p_{i,j}` the j-th
|
||||
value in the probability distribution function, :math:`c_{i,j}` the j-th value
|
||||
in the cumulative distribution function, and :math:`E_{i,j}` the j-th outgoing
|
||||
energy.
|
||||
In a continuous tabular distribution, a tabulated energy distribution is
|
||||
provided for each of a set of incoming energies. While the representation itself
|
||||
is simple, the complexity lies in how one interpolates between incident as well
|
||||
as outgoing energies on such a table. If one performs simple interpolation
|
||||
between tables for neighboring incident energies, it is possible that the
|
||||
resulting energies would violate laws governing the kinematics, i.e., the
|
||||
outgoing energy may be outside the range of available energy in the reaction.
|
||||
|
||||
We proceed first as we did for ACE Law 1, determining the bounding energies of
|
||||
the particle's incoming energy such that :math:`E_i < E < E_{i+1}` and
|
||||
calculating an interpolation factor :math:`f` with equation
|
||||
:eq:`interpolation-factor`. Next, statistical interpolation is performed to
|
||||
choose between using the outgoing energy distributions corresponding to energy
|
||||
:math:`E_i` and :math:`E_{i+1}`. Let :math:`\ell` be the chosen table where
|
||||
:math:`\ell = i` if :math:`\xi_1 > f` and :math:`\ell = i + 1` otherwise, and
|
||||
:math:`\xi_1` is a random number. Then, we sample an outgoing energy bin
|
||||
To avoid this situation, the accepted practice is to use a process known as
|
||||
scaled interpolation [Doyas]_. First, we find the tabulated incident energies
|
||||
which bound the actual incoming energy of the particle, i.e., find :math:`i`
|
||||
such that :math:`E_i < E < E_{i+1}` and calculate the interpolation factor
|
||||
:math:`f` via :eq:`interpolation-factor`. Then, we interpolate between the
|
||||
minimum and maximum energies of the outgoing energy distributions corresponding
|
||||
to :math:`E_i` and :math:`E_{i+1}`:
|
||||
|
||||
.. math::
|
||||
:label: continuous-minmax
|
||||
|
||||
E_{min} = E_{i,1} + f ( E_{i+1,1} - E_{i,1} ) \\
|
||||
E_{max} = E_{i,M} + f ( E_{i+1,M} - E_{i,M} )
|
||||
|
||||
where :math:`E_{min}` and :math:`E_{max}` are the minimum and maximum outgoing
|
||||
energies of a scaled distribution, :math:`E_{i,j}` is the j-th outgoing energy
|
||||
corresponding to the incoming energy :math:`E_i`, and :math:`M` is the number of
|
||||
outgoing energy bins.
|
||||
|
||||
Next, statistical interpolation is performed to choose between using the
|
||||
outgoing energy distributions corresponding to energy :math:`E_i` and
|
||||
:math:`E_{i+1}`. Let :math:`\ell` be the chosen table where :math:`\ell = i` if
|
||||
:math:`\xi_1 > f` and :math:`\ell = i + 1` otherwise, and :math:`\xi_1` is a
|
||||
random number. For each incoming neutron energy :math:`E_i`, let us call
|
||||
:math:`p_{i,j}` the j-th value in the probability distribution function,
|
||||
:math:`c_{i,j}` the j-th value in the cumulative distribution function, and
|
||||
:math:`E_{i,j}` the j-th outgoing energy. We then sample an outgoing energy bin
|
||||
:math:`j` using the cumulative distribution function:
|
||||
|
||||
.. math::
|
||||
:label: ace-law-4-sample-cdf
|
||||
:label: continuous-sample-cdf
|
||||
|
||||
c_{\ell,j} < \xi_2 < c_{\ell,j+1}
|
||||
|
||||
|
|
@ -612,22 +633,22 @@ If linear-linear interpolation is to be used, the outgoing energy on the
|
|||
\right ).
|
||||
|
||||
Since this outgoing energy may violate reaction kinematics, we then scale it to
|
||||
minimum and maximum energies interpolated between the neighboring outgoing
|
||||
energy distributions to get the final outgoing energy:
|
||||
minimum and maximum energies calculated in equation :eq:`continuous-minmax` to
|
||||
get the final outgoing energy:
|
||||
|
||||
.. math::
|
||||
:label: ace-law-4-energy
|
||||
:label: continuous-eout
|
||||
|
||||
E' = E_{min} + \frac{\hat{E} - E_{\ell,1}}{E_{\ell,M} - E_{\ell,1}}
|
||||
(E_{max} - E_{min})
|
||||
|
||||
where :math:`E_{min}` and :math:`E_{max}` are defined the same as in equation
|
||||
:eq:`ace-law-1-minmax`.
|
||||
:eq:`continuous-minmax`.
|
||||
|
||||
.. _maxwell:
|
||||
|
||||
ACE Law 7 - Maxwell Fission Spectrum
|
||||
++++++++++++++++++++++++++++++++++++
|
||||
Maxwell Fission Spectrum
|
||||
^^^^^^^^^^^^^^^^^^^^^^^^
|
||||
|
||||
One representation of the secondary energies for neutrons from fission is the
|
||||
so-called Maxwell spectrum. A probability distribution for the Maxwell spectrum
|
||||
|
|
@ -640,7 +661,7 @@ can be written in the form
|
|||
|
||||
where :math:`E` is the incoming energy of the neutron and :math:`T` is the
|
||||
so-called nuclear temperature, which is a function of the incoming energy of the
|
||||
neutron. The ACE format contains a list of nuclear temperatures versus incoming
|
||||
neutron. The ENDF format contains a list of nuclear temperatures versus incoming
|
||||
energies. The nuclear temperature is interpolated between neighboring incoming
|
||||
energies using a specified interpolation law. Once the temperature :math:`T` is
|
||||
determined, we then calculate a candidate outgoing energy based on rule C64 in
|
||||
|
|
@ -660,12 +681,12 @@ interval. The outgoing energy is only accepted if
|
|||
|
||||
0 \le E' \le E - U
|
||||
|
||||
where :math:`U` is called the restriction energy and is specified on the ACE
|
||||
table. If the outgoing energy is rejected, it is resampled using equation
|
||||
where :math:`U` is called the restriction energy and is specified in the ENDF
|
||||
data. If the outgoing energy is rejected, it is resampled using equation
|
||||
:eq:`maxwell-E-candidate`.
|
||||
|
||||
ACE Law 9 - Evaporation Spectrum
|
||||
++++++++++++++++++++++++++++++++
|
||||
Evaporation Spectrum
|
||||
^^^^^^^^^^^^^^^^^^^^
|
||||
|
||||
Evaporation spectra are primarily used in compound nucleus processes where a
|
||||
secondary particle can "evaporate" from the compound nucleus if it has
|
||||
|
|
@ -679,7 +700,7 @@ be written in the form
|
|||
|
||||
where :math:`E` is the incoming energy of the neutron and :math:`T` is the
|
||||
nuclear temperature, which is a function of the incoming energy of the
|
||||
neutron. The ACE format contains a list of nuclear temperatures versus incoming
|
||||
neutron. The ENDF format contains a list of nuclear temperatures versus incoming
|
||||
energies. The nuclear temperature is interpolated between neighboring incoming
|
||||
energies using a specified interpolation law. Once the temperature :math:`T` is
|
||||
determined, we then calculate a candidate outgoing energy based on the algorithm
|
||||
|
|
@ -697,11 +718,11 @@ energy as in equation :eq:`maxwell-restriction`. This algorithm has a much
|
|||
higher rejection efficiency than the standard technique, i.e. rule C45 in the
|
||||
`Monte Carlo Sampler`_.
|
||||
|
||||
ACE Law 11 - Energy-Dependent Watt Spectrum
|
||||
+++++++++++++++++++++++++++++++++++++++++++
|
||||
Energy-Dependent Watt Spectrum
|
||||
^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^
|
||||
|
||||
The probability distribution for a Watt fission spectrum can be written in the
|
||||
form
|
||||
The probability distribution for a [Watt]_ fission spectrum can be written in
|
||||
the form
|
||||
|
||||
.. math::
|
||||
:label: watt-spectrum
|
||||
|
|
@ -725,29 +746,37 @@ where :math:`\xi` is a random number sampled on the interval :math:`[0,1)`. The
|
|||
outgoing energy is only accepted according to a specified restriction energy
|
||||
:math:`U` as defined in equation :eq:`maxwell-restriction`.
|
||||
|
||||
This algorithm can be found in Forrest Brown's lectures_ on Monte Carlo methods
|
||||
and is an unpublished sampling scheme based on the original Watt spectrum
|
||||
derivation [Watt]_.
|
||||
A derivation of the algorithm described here can be found in a paper by Romano_.
|
||||
|
||||
ACE Law 44 - Kalbach-Mann Correlated Scattering
|
||||
+++++++++++++++++++++++++++++++++++++++++++++++
|
||||
Product Angle-Energy Distributions
|
||||
----------------------------------
|
||||
|
||||
This law is very similar to ACE Law 4 except now the outgoing angle of the
|
||||
neutron is correlated to the outgoing energy and is not sampled from a separate
|
||||
distribution. For each incident neutron energy :math:`E_i` tabulated, there is
|
||||
an array of precompound factors :math:`R_{i,j}` and angular distribution slopes
|
||||
:math:`A_{i,j}` corresponding to each outgoing energy bin :math:`j` in addition
|
||||
to the outgoing energies and distribution functions as in ACE Law 4.
|
||||
If the secondary distribution for a product was given in file 6 in ENDF, the
|
||||
angle and energy are correlated with one another and cannot be sampled
|
||||
separately. Several representations exist in ENDF/ACE for correlated
|
||||
angle-energy distributions.
|
||||
|
||||
Kalbach-Mann Correlated Scattering
|
||||
++++++++++++++++++++++++++++++++++
|
||||
|
||||
This law is very similar to the uncorrelated continuous tabular energy
|
||||
distribution except now the outgoing angle of the neutron is correlated to the
|
||||
outgoing energy and is not sampled from a separate distribution. For each
|
||||
incident neutron energy :math:`E_i` tabulated, there is an array of precompound
|
||||
factors :math:`R_{i,j}` and angular distribution slopes :math:`A_{i,j}`
|
||||
corresponding to each outgoing energy bin :math:`j` in addition to the outgoing
|
||||
energies and distribution functions as in :ref:`continuous-tabular`.
|
||||
|
||||
The calculation of the outgoing energy of the neutron proceeds exactly the same
|
||||
as in the algorithm described in :ref:`ace-law-4`. In that algorithm, we found
|
||||
an interpolation factor :math:`f`, statistically sampled an incoming energy bin
|
||||
:math:`\ell`, and sampled an outgoing energy bin :math:`j` based on the
|
||||
tabulated cumulative distribution function. Once the outgoing energy has been
|
||||
determined with equation :eq:`ace-law-4-energy`, we then need to calculate the
|
||||
outgoing angle based on the tabulated Kalbach-Mann parameters. These parameters
|
||||
themselves are subject to either histogram or linear-linear interpolation on the
|
||||
outgoing energy grid. For histogram interpolation, the parameters are
|
||||
as in the algorithm described in :ref:`continuous-tabular`. In that algorithm,
|
||||
we found an interpolation factor :math:`f`, statistically sampled an incoming
|
||||
energy bin :math:`\ell`, and sampled an outgoing energy bin :math:`j` based on
|
||||
the tabulated cumulative distribution function. Once the outgoing energy has
|
||||
been determined with equation :eq:`continuous-eout`, we then need to calculate
|
||||
the outgoing angle based on the tabulated Kalbach-Mann parameters. These
|
||||
parameters themselves are subject to either histogram or linear-linear
|
||||
interpolation on the outgoing energy grid. For histogram interpolation, the
|
||||
parameters are
|
||||
|
||||
.. math::
|
||||
:label: KM-parameters-histogram
|
||||
|
|
@ -793,52 +822,55 @@ outgoing angle is
|
|||
|
||||
\mu = \frac{1}{A} \ln \left ( \xi_4 e^A + (1 - \xi_4) e^{-A} \right ).
|
||||
|
||||
.. _ace-law-61:
|
||||
.. _correlated-energy-angle:
|
||||
|
||||
ACE Law 61 - Correlated Energy and Angle Distribution
|
||||
+++++++++++++++++++++++++++++++++++++++++++++++++++++
|
||||
Correlated Energy and Angle Distribution
|
||||
++++++++++++++++++++++++++++++++++++++++
|
||||
|
||||
This law is very similar to ACE Law 44 in the sense that the outgoing angle of
|
||||
the neutron is correlated to the outgoing energy and is not sampled from a
|
||||
separate distribution. In this case though, rather than being determined from an
|
||||
analytical distribution function, the cosine of the scattering angle is
|
||||
determined from a tabulated distribution. For each incident energy :math:`i` and
|
||||
outgoing energy :math:`j`, there is a tabulated angular distribution.
|
||||
This distribution is very similar to a Kalbach-Mann distribution in the sense
|
||||
that the outgoing angle of the neutron is correlated to the outgoing energy and
|
||||
is not sampled from a separate distribution. In this case though, rather than
|
||||
being determined from an analytical distribution function, the cosine of the
|
||||
scattering angle is determined from a tabulated distribution. For each incident
|
||||
energy :math:`i` and outgoing energy :math:`j`, there is a tabulated angular
|
||||
distribution.
|
||||
|
||||
The calculation of the outgoing energy of the neutron proceeds exactly the same
|
||||
as in the algorithm described in :ref:`ace-law-4`. In that algorithm, we found
|
||||
an interpolation factor :math:`f`, statistically sampled an incoming energy bin
|
||||
:math:`\ell`, and sampled an outgoing energy bin :math:`j` based on the
|
||||
tabulated cumulative distribution function. Once the outgoing energy has been
|
||||
determined with equation :eq:`ace-law-4-energy`, we then need to decide which
|
||||
angular distribution to use. If histogram interpolation was used on the outgoing
|
||||
energy bins, then we use the angular distribution corresponding to incoming
|
||||
energy bin :math:`\ell` and outgoing energy bin :math:`j`. If linear-linear
|
||||
interpolation was used on the outgoing energy bins, then we use the whichever
|
||||
angular distribution was closer to the sampled value of the cumulative
|
||||
distribution function for the outgoing energy. The actual algorithm used to
|
||||
sample the chosen tabular angular distribution has been previously described in
|
||||
:ref:`angle-tabular`.
|
||||
as in the algorithm described in :ref:`continuous-tabular`. In that algorithm,
|
||||
we found an interpolation factor :math:`f`, statistically sampled an incoming
|
||||
energy bin :math:`\ell`, and sampled an outgoing energy bin :math:`j` based on
|
||||
the tabulated cumulative distribution function. Once the outgoing energy has
|
||||
been determined with equation :eq:`continuous-eout`, we then need to decide
|
||||
which angular distribution to use. If histogram interpolation was used on the
|
||||
outgoing energy bins, then we use the angular distribution corresponding to
|
||||
incoming energy bin :math:`\ell` and outgoing energy bin :math:`j`. If
|
||||
linear-linear interpolation was used on the outgoing energy bins, then we use
|
||||
the whichever angular distribution was closer to the sampled value of the
|
||||
cumulative distribution function for the outgoing energy. The actual algorithm
|
||||
used to sample the chosen tabular angular distribution has been previously
|
||||
described in :ref:`angle-tabular`.
|
||||
|
||||
ACE Law 66 - N-Body Phase Space Distribution
|
||||
++++++++++++++++++++++++++++++++++++++++++++
|
||||
N-Body Phase Space Distribution
|
||||
+++++++++++++++++++++++++++++++
|
||||
|
||||
Reactions in which there are more than two products of similar masses are
|
||||
sometimes best treated by using what's known as an N-body phase
|
||||
distribution. This distribution has the following probability density function
|
||||
for outgoing energy of the :math:`i`-th particle in the center-of-mass system:
|
||||
for outgoing energy and angle of the :math:`i`-th particle in the center-of-mass
|
||||
system:
|
||||
|
||||
.. math::
|
||||
:label: n-body-pdf
|
||||
|
||||
p_i(E') dE' = C_n \sqrt{E'} (E_i^{max} - E')^{(3n/2) - 4} dE'
|
||||
p_i(\mu, E') dE' d\mu = C_n \sqrt{E'} (E_i^{max} - E')^{(3n/2) - 4} dE' d\mu
|
||||
|
||||
where :math:`n` is the number of outgoing particles, :math:`C_n` is a
|
||||
normalization constant, :math:`E_i^{max}` is the maximum center-of-mass energy
|
||||
for particle :math:`i`, and :math:`E'` is the outgoing energy. The algorithm for
|
||||
sampling the outgoing energy is based on algorithms R28, C45, and C64 in the
|
||||
`Monte Carlo Sampler`_. First we calculate the maximum energy in the
|
||||
center-of-mass using the following equation:
|
||||
for particle :math:`i`, and :math:`E'` is the outgoing energy. We see in
|
||||
equation :eq:`n-body-pdf` that the angle is simply isotropic in the
|
||||
center-of-mass system. The algorithm for sampling the outgoing energy is based
|
||||
on algorithms R28, C45, and C64 in the `Monte Carlo Sampler`_. First we
|
||||
calculate the maximum energy in the center-of-mass using the following equation:
|
||||
|
||||
.. math::
|
||||
:label: n-body-emax
|
||||
|
|
@ -881,7 +913,7 @@ distribution. First, the documentation (and code) for MCNP5-1.60 has a mistake
|
|||
in the algorithm for :math:`n = 4`. That being said, there are no existing
|
||||
nuclear data evaluations which use an N-body phase space distribution with
|
||||
:math:`n = 4`, so the error would not affect any calculations. In the
|
||||
ENDF/B-VII.0 nuclear data evaluation, only one reaction uses an N-body phase
|
||||
ENDF/B-VII.1 nuclear data evaluation, only one reaction uses an N-body phase
|
||||
space distribution at all, the :math:`(n,2n)` reaction with H-2.
|
||||
|
||||
.. _transform-coordinates:
|
||||
|
|
@ -890,6 +922,9 @@ space distribution at all, the :math:`(n,2n)` reaction with H-2.
|
|||
Transforming a Particle's Coordinates
|
||||
-------------------------------------
|
||||
|
||||
Since all the multi-group data exists in the laboratory frame of reference, this
|
||||
section does not apply to the multi-group mode.
|
||||
|
||||
Once the cosine of the scattering angle :math:`\mu` has been sampled either from
|
||||
a angle distribution or a correlated angle-energy distribution, we are still
|
||||
left with the task of transforming the particle's coordinates. If the outgoing
|
||||
|
|
@ -941,6 +976,9 @@ the post-collision direction is calculated as
|
|||
Effect of Thermal Motion on Cross Sections
|
||||
------------------------------------------
|
||||
|
||||
Since all the multi-group data should be generated with thermal scattering
|
||||
treatments already, this section does not apply to the multi-group mode.
|
||||
|
||||
When a neutron scatters off of a nucleus, it may often be assumed that the
|
||||
target nucleus is at rest. However, the target nucleus will have motion
|
||||
associated with its thermal vibration, even at absolute zero (This is due to the
|
||||
|
|
@ -1272,6 +1310,8 @@ described fully in `Walsh et al.`_
|
|||
|sab| Tables
|
||||
------------
|
||||
|
||||
Note that |sab| tables are only applicable to continuous-energy transport.
|
||||
|
||||
For neutrons with thermal energies, generally less than 4 eV, the kinematics of
|
||||
scattering can be affected by chemical binding and crystalline effects of the
|
||||
target molecule. If these effects are not accounted for in a simulation, the
|
||||
|
|
@ -1439,16 +1479,16 @@ accordingly.
|
|||
Continuous Outgoing Energies
|
||||
++++++++++++++++++++++++++++
|
||||
|
||||
If the thermal data was processed with :math:`iwt=2` in NJOY, then the
|
||||
outgoing energy spectra is represented by a continuous outgoing energy spectra
|
||||
in tabular form with linear-linear interpolation. The sampling of the outgoing
|
||||
energy portion of this format is very similar to :ref:`ACE Law 61<ace-law-61>`,
|
||||
but the sampling of the correlated angle is performed as it was in the other
|
||||
two representations discussed in this sub-section. In the Law 61 algorithm,
|
||||
we found an interpolation factor :math:`f`, statistically sampled an incoming
|
||||
If the thermal data was processed with :math:`iwt=2` in NJOY, then the outgoing
|
||||
energy spectra is represented by a continuous outgoing energy spectra in tabular
|
||||
form with linear-linear interpolation. The sampling of the outgoing energy
|
||||
portion of this format is very similar to :ref:`correlated-energy-angle`, but
|
||||
the sampling of the correlated angle is performed as it was in the other two
|
||||
representations discussed in this sub-section. In the Law 61 algorithm, we
|
||||
found an interpolation factor :math:`f`, statistically sampled an incoming
|
||||
energy bin :math:`\ell`, and sampled an outgoing energy bin :math:`j` based on
|
||||
the tabulated cumulative distribution function. Once the outgoing energy has
|
||||
been determined with equation :eq:`ace-law-4-energy`, we then need to decide
|
||||
been determined with equation :eq:`continuous-eout`, we then need to decide
|
||||
which angular distribution data to use. Like the linear-linear interpolation
|
||||
case in Law 61, the angular distribution closest to the sampled value of the
|
||||
cumulative distribution function for the outgoing energy is utilized. The
|
||||
|
|
@ -1461,6 +1501,9 @@ actual algorithm utilized to sample the outgoing angle is shown in equation
|
|||
Unresolved Resonance Region Probability Tables
|
||||
----------------------------------------------
|
||||
|
||||
Note that unresolved resonance treatments are only applicable to
|
||||
continuous-energy transport.
|
||||
|
||||
In the unresolved resonance energy range, resonances may be so closely spaced
|
||||
that it is not possible for experimental measurements to resolve all
|
||||
resonances. To properly account for self-shielding in this energy range, OpenMC
|
||||
|
|
@ -1632,6 +1675,8 @@ another.
|
|||
|
||||
.. _MC21: http://www.osti.gov/bridge/servlets/purl/903083-HT5p1o/903083.pdf
|
||||
|
||||
.. _Romano: http://dx.doi.org/10.1016/j.cpc.2014.11.001
|
||||
|
||||
.. _Sutton and Brown: http://www.osti.gov/bridge/product.biblio.jsp?osti_id=307911
|
||||
|
||||
.. _lectures: https://laws.lanl.gov/vhosts/mcnp.lanl.gov/pdf_files/la-ur-05-4983.pdf
|
||||
|
|
|
|||
|
|
@ -4,6 +4,10 @@
|
|||
Tallies
|
||||
=======
|
||||
|
||||
Note that the methods discussed in this section are written specifically for
|
||||
continuous-energy mode but equivalent apply to the multi-group mode if the
|
||||
particle's energy is replaced with the particle's group
|
||||
|
||||
------------------
|
||||
Filters and Scores
|
||||
------------------
|
||||
|
|
@ -32,8 +36,9 @@ OpenMC: flux, total reaction rate, scattering reaction rate, neutron production
|
|||
from scattering, higher scattering moments, :math:`(n,xn)` reaction rates,
|
||||
absorption reaction rate, fission reaction rate, neutron production rate from
|
||||
fission, and surface currents. The following variables can be used as filters:
|
||||
universe, material, cell, birth cell, surface, mesh, pre-collision energy, and
|
||||
post-collision energy.
|
||||
universe, material, cell, birth cell, surface, mesh, pre-collision energy,
|
||||
post-collision energy, polar angle, azimuthal angle, and the cosine of the
|
||||
change-in-angle due to a scattering event.
|
||||
|
||||
With filters for pre- and post-collision energy and scoring functions for
|
||||
scattering and fission production, it is possible to use OpenMC to generate
|
||||
|
|
@ -55,9 +60,9 @@ be scored to for each value of the filter variable. If a particle is in cell
|
|||
:math:`n`, the mapping would identify what tally/bin combinations specify cell
|
||||
:math:`n` for the cell filter variable. In this manner, it is not necessary to
|
||||
check the phase space variables against each tally. Note that this technique
|
||||
only applies to discrete filter variables and cannot be applied to energy
|
||||
bins. For energy filters, it is necessary to perform a binary search on the
|
||||
specified energy grid.
|
||||
only applies to discrete filter variables and cannot be applied to energy,
|
||||
angle, or change-in-angle bins. For these filters, it is necessary to perform
|
||||
a binary search on the specified energy grid.
|
||||
|
||||
-----------------------------------------
|
||||
Volume-Integrated Flux and Reaction Rates
|
||||
|
|
@ -196,8 +201,9 @@ One important fact to take into consideration is that the use of a track-length
|
|||
estimator precludes us from using any filter that requires knowledge of the
|
||||
particle's state following a collision because by definition, it will not have
|
||||
had a collision at every event. Thus, for tallies with outgoing-energy filters
|
||||
(which require the post-collision energy) or for tallies of scattering moments
|
||||
(which require the scattering cosine), we must use an analog estimator.
|
||||
(which require the post-collision energy), scattering change-in-angle filters,
|
||||
or for tallies of scattering moments (which require the scattering cosine of
|
||||
the change-in-angle), we must use an analog estimator.
|
||||
|
||||
.. TODO: Add description of surface current tallies
|
||||
|
||||
|
|
@ -430,7 +436,7 @@ analytically. For one degree of freedom, the t-distribution becomes a standard
|
|||
.. math::
|
||||
:label: cauchy-cdf
|
||||
|
||||
c(x) = \frac{1}{\pi} \arctan x + \frac{1}{2}.
|
||||
c(x) = \frac{1}{\pi} \arctan x + \frac{1}{2}.
|
||||
|
||||
Thus, inverting the cumulative distribution function, we find the :math:`x`
|
||||
percentile of the standard Cauchy distribution to be
|
||||
|
|
|
|||
|
|
@ -53,6 +53,16 @@ Benchmarking
|
|||
Coupling and Multi-physics
|
||||
--------------------------
|
||||
|
||||
- Matthew Ellis, Benoit Forget, Kord Smith, and Derek Gaston, "Continuous
|
||||
Temperature Representation in Coupled OpenMC/MOOSE Simulations," *Proc. PHYSOR
|
||||
2016*, Sun Valley, Idaho, May 1-5, 2016.
|
||||
|
||||
- Antonios G. Mylonakis, Melpomeni Varvayanni, and Nicolas Catsaros,
|
||||
"Investigating a Matrix-free, Newton-based, Neutron-Monte
|
||||
Carlo/Thermal-Hydraulic Coupling Scheme", *Proc. Int. Conf. Nuclear Energy for
|
||||
New Europe*, Portoroz, Slovenia, Sep .14-17
|
||||
(2015). `<https://www.researchgate.net/publication/282001032>`_
|
||||
|
||||
- Matt Ellis, Benoit Forget, Kord Smith, and Derek Gaston, "Preliminary coupling
|
||||
of the Monte Carlo code OpenMC and the Multiphysics Object-Oriented Simulation
|
||||
Environment (MOOSE) for analyzing Doppler feedback in Monte Carlo
|
||||
|
|
@ -80,8 +90,17 @@ Geometry
|
|||
Miscellaneous
|
||||
-------------
|
||||
|
||||
- Yunzhao Li, Qingming He, Liangzhi Cao, Hongchun Wu, and Tiejun Zu, "Resonance
|
||||
Elastic Scattering and Interference Effects Treatments in Subgroup Method,"
|
||||
*Nucl. Eng. Tech.*, **48**, 339-350
|
||||
(2016). `<http://dx.doi.org/10.1016/j.net.2015.12.015>`_
|
||||
|
||||
- William Boyd, Sterling Harper, and Paul K. Romano, "Equipping OpenMC for the
|
||||
big data era," Accepted, *PHYSOR 2016*, Sun Valley, Idaho, May 1-5, 2016.
|
||||
big data era," *Proc. PHYSOR*, Sun Valley, Idaho, May 1-5, 2016.
|
||||
|
||||
- Michal Kostal, Vojtech Rypar, Jan Milcak, Vlastimil Juricek, Evzen Losa,
|
||||
Benoit Forget, and Sterling Harper, *Ann. Nucl. Energy*, **87**, 601-611
|
||||
(2016). `<http://dx.doi.org/10.1016/j.anucene.2015.10.010>`_
|
||||
|
||||
- Qicang Shen, William Boyd, Benoit Forget, and Kord Smith, "Tally precision
|
||||
triggers for the OpenMC Monte Carlo code," *Trans. Am. Nucl. Soc.*, **112**,
|
||||
|
|
@ -95,6 +114,11 @@ Miscellaneous
|
|||
Multi-group Cross Section Generation
|
||||
------------------------------------
|
||||
|
||||
- Zhaoyuan Liu, Kord Smith, and Benoit Forget, "A Cumulative Migration Method
|
||||
for Computing Rigorous Transport Cross Sections and Diffusion Coefficients for
|
||||
LWR Lattices with Monte Carlo," *Proc. PHYSOR*, Sun Valley, Idaho, May
|
||||
1-5, 2016.
|
||||
|
||||
- Adam G. Nelson and William R. Martin, "Improved Monte Carlo tallying of
|
||||
multi-group scattering moments using the NDPP code," *Trans. Am. Nucl. Soc.*,
|
||||
**113**, 645-648 (2015)
|
||||
|
|
@ -108,18 +132,43 @@ Multi-group Cross Section Generation
|
|||
Computational Methods Applied to Nuclear Science and Engineering*, Sun Valley,
|
||||
Idaho, May 5--9 (2013).
|
||||
|
||||
------------
|
||||
Nuclear Data
|
||||
------------
|
||||
|
||||
------------------
|
||||
Doppler Broadening
|
||||
------------------
|
||||
|
||||
- Colin Josey, Pablo Ducru, Benoit Forget, and Kord Smith, "Windowed multipole
|
||||
for cross section Doppler broadening," *J. Comput. Phys.*, In Press
|
||||
for cross section Doppler broadening," *J. Comput. Phys.*, **307**, 715-727
|
||||
(2016). `<http://dx.doi.org/10.1016/jcp.2015.08.013>`_
|
||||
|
||||
- Jonathan A. Walsh, Benoit Forget, Kord S. Smith, and Forrest B. Brown,
|
||||
"On-the-fly Doppler Broadening of Unresolved Resonance Region Cross Sections
|
||||
via Probability Band Interpolation," *Proc. PHYSOR*, Sun Valley, Idaho, May
|
||||
1-5, 2016.
|
||||
|
||||
- Colin Josey, Benoit Forget, and Kord Smith, "Windowed multipole sensitivity to
|
||||
target accuracy of the optimization procedure," *J. Nucl. Sci. Technol.*,
|
||||
**52**, 987-992 (2015). `<http://dx.doi.org/10.1080/00223131.2015.1035353>`_
|
||||
|
||||
- Paul K. Romano and Timothy H. Trumbull, "Comparison of algorithms for Doppler
|
||||
broadening pointwise tabulated cross sections," *Ann. Nucl. Energy*, **75**,
|
||||
358--364 (2015). `<http://dx.doi.org/10.1016/j.anucene.2014.08.046>`_
|
||||
|
||||
- Tuomas Viitanen, Jaakko Leppanen, and Benoit Forget, "Target motion sampling
|
||||
temperature treatment technique with track-length esimators in OpenMC --
|
||||
Preliminary results," *Proc. PHYSOR*, Kyoto, Japan, Sep. 28--Oct. 3 (2014).
|
||||
|
||||
- Benoit Forget, Sheng Xu, and Kord Smith, "Direct Doppler broadening in Monte
|
||||
Carlo simulations using the multipole representation," *Ann. Nucl. Energy*,
|
||||
**64**, 78--85 (2014). `<http://dx.doi.org/10.1016/j.anucene.2013.09.043>`_
|
||||
|
||||
------------
|
||||
Nuclear Data
|
||||
------------
|
||||
|
||||
- Paul K. Romano and Sterling M. Harper, "Nuclear data processing capabilities
|
||||
in OpenMC", *Proc. Nuclear Data*, Sep. 11-16, 2016.
|
||||
|
||||
- Jonathan A. Walsh, Paul K. Romano, Benoit Forget, and Kord S. Smith,
|
||||
"Optimizations of the energy grid search algorithm in continuous-energy Monte
|
||||
Carlo particle transport codes", *Comput. Phys. Commun.*, **196**, 134-142
|
||||
|
|
@ -139,29 +188,17 @@ Nuclear Data
|
|||
performance analysis for varying cross section parameter regimes,"
|
||||
*Proc. Joint Int. Conf. M&C+SNA+MC*, Nashville, Tennessee, Apr. 19--23 (2015).
|
||||
|
||||
- Paul K. Romano and Timothy H. Trumbull, "Comparison of algorithms for Doppler
|
||||
broadening pointwise tabulated cross sections," *Ann. Nucl. Energy*, **75**,
|
||||
358--364 (2015). `<http://dx.doi.org/10.1016/j.anucene.2014.08.046>`_
|
||||
|
||||
- Tuomas Viitanen, Jaakko Leppanen, and Benoit Forget, "Target motion sampling
|
||||
temperature treatment technique with track-length esimators in OpenMC --
|
||||
Preliminary results," *Proc. PHYSOR*, Kyoto, Japan, Sep. 28--Oct. 3 (2014).
|
||||
|
||||
- Jonathan A. Walsh, Benoit Forget, and Kord S. Smith, "Accelerated sampling
|
||||
of the free gas resonance elastic scattering kernel," *Ann. Nucl. Energy*,
|
||||
**69**, 116--124 (2014). `<http://dx.doi.org/10.1016/j.anucene.2014.01.017>`_
|
||||
|
||||
- Benoit Forget, Sheng Xu, and Kord Smith, "Direct Doppler broadening in Monte
|
||||
Carlo simulations using the multipole representation," *Ann. Nucl. Energy*,
|
||||
**64**, 78--85 (2014). `<http://dx.doi.org/10.1016/j.anucene.2013.09.043>`_
|
||||
|
||||
-----------
|
||||
Parallelism
|
||||
-----------
|
||||
|
||||
- Paul K. Romano, John R. Tramm, and Andrew R. Siegel, "Efficacy of hardware
|
||||
threading for Monte Carlo particle transport calculations on multi- and
|
||||
many-core systems," Accepted, *PHYSOR 2016*, Sun Valley, Idaho, May 1-5, 2016.
|
||||
many-core systems," *PHYSOR 2016*, Sun Valley, Idaho, May 1-5, 2016.
|
||||
|
||||
- David Ozog, Allen D. Malony, and Andrew R. Siegel, "A performance analysis of
|
||||
SIMD algorithms for Monte Carlo simulations of nuclear reactor cores,"
|
||||
|
|
@ -228,3 +265,11 @@ Parallelism
|
|||
- Paul K. Romano and Benoit Forget, "Parallel Fission Bank Algorithms in Monte
|
||||
Carlo Criticality Calculations," *Nucl. Sci. Eng.*, **170**, 125--135
|
||||
(2012). `<http://hdl.handle.net/1721.1/73569>`_
|
||||
|
||||
---------
|
||||
Depletion
|
||||
---------
|
||||
|
||||
- Kai Huang, Hongchun Wu, Yunzhao Li, and Liangzhi Cao, "Generalized depletion
|
||||
chain simplification based of significance analysis," *Proc. PHYSOR*, Sun
|
||||
Valley, Idaho, May 1-5, 2016.
|
||||
|
|
|
|||
BIN
docs/source/pythonapi/examples/images/mdgxs.png
Normal file
BIN
docs/source/pythonapi/examples/images/mdgxs.png
Normal file
Binary file not shown.
|
After Width: | Height: | Size: 23 KiB |
1559
docs/source/pythonapi/examples/mdgxs-part-i.ipynb
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1559
docs/source/pythonapi/examples/mdgxs-part-i.ipynb
Normal file
File diff suppressed because one or more lines are too long
13
docs/source/pythonapi/examples/mdgxs-part-i.rst
Normal file
13
docs/source/pythonapi/examples/mdgxs-part-i.rst
Normal file
|
|
@ -0,0 +1,13 @@
|
|||
.. _notebook_mdgxs_part_i:
|
||||
|
||||
==========================
|
||||
MDGXS Part I: Introduction
|
||||
==========================
|
||||
|
||||
.. only:: html
|
||||
|
||||
.. notebook:: mdgxs-part-i.ipynb
|
||||
|
||||
.. only:: latex
|
||||
|
||||
IPython notebooks must be viewed in the online HTML documentation.
|
||||
1328
docs/source/pythonapi/examples/mdgxs-part-ii.ipynb
Normal file
1328
docs/source/pythonapi/examples/mdgxs-part-ii.ipynb
Normal file
File diff suppressed because one or more lines are too long
13
docs/source/pythonapi/examples/mdgxs-part-ii.rst
Normal file
13
docs/source/pythonapi/examples/mdgxs-part-ii.rst
Normal file
|
|
@ -0,0 +1,13 @@
|
|||
.. _notebook_mdgxs_part_ii:
|
||||
|
||||
================================
|
||||
MDGXS Part II: Advanced Features
|
||||
================================
|
||||
|
||||
.. only:: html
|
||||
|
||||
.. notebook:: mdgxs-part-ii.ipynb
|
||||
|
||||
.. only:: latex
|
||||
|
||||
IPython notebooks must be viewed in the online HTML documentation.
|
||||
|
|
@ -165,11 +165,11 @@
|
|||
"outputs": [],
|
||||
"source": [
|
||||
"# Instantiate some Nuclides\n",
|
||||
"h1 = openmc.Nuclide('H-1')\n",
|
||||
"o16 = openmc.Nuclide('O-16')\n",
|
||||
"u235 = openmc.Nuclide('U-235')\n",
|
||||
"u238 = openmc.Nuclide('U-238')\n",
|
||||
"zr90 = openmc.Nuclide('Zr-90')"
|
||||
"h1 = openmc.Nuclide('H1')\n",
|
||||
"o16 = openmc.Nuclide('O16')\n",
|
||||
"u235 = openmc.Nuclide('U235')\n",
|
||||
"u238 = openmc.Nuclide('U238')\n",
|
||||
"zr90 = openmc.Nuclide('Zr90')"
|
||||
]
|
||||
},
|
||||
{
|
||||
|
|
@ -214,7 +214,6 @@
|
|||
"source": [
|
||||
"# Instantiate a Materials collection and export to XML\n",
|
||||
"materials_file = openmc.Materials([inf_medium])\n",
|
||||
"materials_file.default_xs = '71c'\n",
|
||||
"materials_file.export_to_xml()"
|
||||
]
|
||||
},
|
||||
|
|
@ -383,6 +382,9 @@
|
|||
"* `ScatterMatrixXS`\n",
|
||||
"* `NuScatterMatrixXS`\n",
|
||||
"* `Chi`\n",
|
||||
"* `ChiPrompt`\n",
|
||||
"* `InverseVelocity`\n",
|
||||
"* `PromptNuFissionXS`\n",
|
||||
"\n",
|
||||
"These classes provide us with an interface to generate the tally inputs as well as perform post-processing of OpenMC's tally data to compute the respective multi-group cross sections. In this case, let's create the multi-group total, absorption and scattering cross sections with our 2-group structure."
|
||||
]
|
||||
|
|
@ -419,24 +421,22 @@
|
|||
"data": {
|
||||
"text/plain": [
|
||||
"OrderedDict([('flux', Tally\n",
|
||||
"\tID =\t10000\n",
|
||||
"\tName =\t\n",
|
||||
"\tFilters =\t\n",
|
||||
" \t\tcell\t[1]\n",
|
||||
" \t\tenergy\t[ 0.00000000e+00 6.25000000e-07 2.00000000e+01]\n",
|
||||
"\tNuclides =\ttotal \n",
|
||||
"\tScores =\t['flux']\n",
|
||||
"\tEstimator =\ttracklength\n",
|
||||
"), ('absorption', Tally\n",
|
||||
"\tID =\t10001\n",
|
||||
"\tName =\t\n",
|
||||
"\tFilters =\t\n",
|
||||
" \t\tcell\t[1]\n",
|
||||
" \t\tenergy\t[ 0.00000000e+00 6.25000000e-07 2.00000000e+01]\n",
|
||||
"\tNuclides =\ttotal \n",
|
||||
"\tScores =\t['absorption']\n",
|
||||
"\tEstimator =\ttracklength\n",
|
||||
")])"
|
||||
" \tID =\t10000\n",
|
||||
" \tName =\t\n",
|
||||
" \tFilters =\t\n",
|
||||
" \t\tcell\t[1]\n",
|
||||
" \t\tenergy\t[ 0.00000000e+00 6.25000000e-07 2.00000000e+01]\n",
|
||||
" \tNuclides =\ttotal \n",
|
||||
" \tScores =\t['flux']\n",
|
||||
" \tEstimator =\ttracklength), ('absorption', Tally\n",
|
||||
" \tID =\t10001\n",
|
||||
" \tName =\t\n",
|
||||
" \tFilters =\t\n",
|
||||
" \t\tcell\t[1]\n",
|
||||
" \t\tenergy\t[ 0.00000000e+00 6.25000000e-07 2.00000000e+01]\n",
|
||||
" \tNuclides =\ttotal \n",
|
||||
" \tScores =\t['absorption']\n",
|
||||
" \tEstimator =\ttracklength)])"
|
||||
]
|
||||
},
|
||||
"execution_count": 13,
|
||||
|
|
@ -498,41 +498,54 @@
|
|||
"output_type": "stream",
|
||||
"text": [
|
||||
"\n",
|
||||
" .d88888b. 888b d888 .d8888b.\n",
|
||||
" d88P\" \"Y88b 8888b d8888 d88P Y88b\n",
|
||||
" 888 888 88888b.d88888 888 888\n",
|
||||
" 888 888 88888b. .d88b. 88888b. 888Y88888P888 888 \n",
|
||||
" 888 888 888 \"88b d8P Y8b 888 \"88b 888 Y888P 888 888 \n",
|
||||
" 888 888 888 888 88888888 888 888 888 Y8P 888 888 888\n",
|
||||
" Y88b. .d88P 888 d88P Y8b. 888 888 888 \" 888 Y88b d88P\n",
|
||||
" \"Y88888P\" 88888P\" \"Y8888 888 888 888 888 \"Y8888P\"\n",
|
||||
"__________________888______________________________________________________\n",
|
||||
" 888\n",
|
||||
" 888\n",
|
||||
" %%%%%%%%%%%%%%%\n",
|
||||
" %%%%%%%%%%%%%%%%%%%%%%%%\n",
|
||||
" %%%%%%%%%%%%%%%%%%%%%%%%%%%%%%\n",
|
||||
" %%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%\n",
|
||||
" %%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%\n",
|
||||
" %%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%\n",
|
||||
" %%%%%%%%%%%%%%%%%%%%%%%%\n",
|
||||
" %%%%%%%%%%%%%%%%%%%%%%%%\n",
|
||||
" ############### %%%%%%%%%%%%%%%%%%%%%%%%\n",
|
||||
" ################## %%%%%%%%%%%%%%%%%%%%%%%\n",
|
||||
" ################### %%%%%%%%%%%%%%%%%%%%%%%\n",
|
||||
" #################### %%%%%%%%%%%%%%%%%%%%%%\n",
|
||||
" ##################### %%%%%%%%%%%%%%%%%%%%%\n",
|
||||
" ###################### %%%%%%%%%%%%%%%%%%%%\n",
|
||||
" ####################### %%%%%%%%%%%%%%%%%%\n",
|
||||
" ####################### %%%%%%%%%%%%%%%%%\n",
|
||||
" ###################### %%%%%%%%%%%%%%%%%\n",
|
||||
" #################### %%%%%%%%%%%%%%%%%\n",
|
||||
" ################# %%%%%%%%%%%%%%%%%\n",
|
||||
" ############### %%%%%%%%%%%%%%%%\n",
|
||||
" ############ %%%%%%%%%%%%%%%\n",
|
||||
" ######## %%%%%%%%%%%%%%\n",
|
||||
" %%%%%%%%%%%\n",
|
||||
"\n",
|
||||
" Copyright: 2011-2016 Massachusetts Institute of Technology\n",
|
||||
" License: http://openmc.readthedocs.io/en/latest/license.html\n",
|
||||
" Version: 0.7.1\n",
|
||||
" Git SHA1: 19feb55e6d5e8350398627f39fb55ee8e2e63011\n",
|
||||
" Date/Time: 2016-05-13 10:19:16\n",
|
||||
" MPI Processes: 1\n",
|
||||
" | The OpenMC Monte Carlo Code\n",
|
||||
" Copyright | 2011-2016 Massachusetts Institute of Technology\n",
|
||||
" License | http://openmc.readthedocs.io/en/latest/license.html\n",
|
||||
" Version | 0.8.0\n",
|
||||
" Git SHA1 | fbebf7bf709fe2fe1813af95bff9b29c0d59312c\n",
|
||||
" Date/Time | 2016-08-31 10:40:13\n",
|
||||
" OpenMP Threads | 4\n",
|
||||
"\n",
|
||||
" ===========================================================================\n",
|
||||
" ========================> INITIALIZATION <=========================\n",
|
||||
" ===========================================================================\n",
|
||||
"\n",
|
||||
" Reading settings XML file...\n",
|
||||
" Reading cross sections XML file...\n",
|
||||
" Reading geometry XML file...\n",
|
||||
" Reading cross sections XML file...\n",
|
||||
" Reading materials XML file...\n",
|
||||
" Reading H1 from /home/romano/openmc/data/nndc_hdf5/H1.h5\n",
|
||||
" Reading O16 from /home/romano/openmc/data/nndc_hdf5/O16.h5\n",
|
||||
" Reading U235 from /home/romano/openmc/data/nndc_hdf5/U235.h5\n",
|
||||
" Reading U238 from /home/romano/openmc/data/nndc_hdf5/U238.h5\n",
|
||||
" Reading Zr90 from /home/romano/openmc/data/nndc_hdf5/Zr90.h5\n",
|
||||
" Maximum neutron transport energy: 20.0000 MeV for H1\n",
|
||||
" Reading tallies XML file...\n",
|
||||
" Building neighboring cells lists for each surface...\n",
|
||||
" Loading ACE cross section table: 1001.71c\n",
|
||||
" Loading ACE cross section table: 8016.71c\n",
|
||||
" Loading ACE cross section table: 92235.71c\n",
|
||||
" Loading ACE cross section table: 92238.71c\n",
|
||||
" Loading ACE cross section table: 40090.71c\n",
|
||||
" Maximum neutron transport energy: 20.0000 MeV for 1001.71c\n",
|
||||
" Initializing source particles...\n",
|
||||
"\n",
|
||||
" ===========================================================================\n",
|
||||
|
|
@ -600,20 +613,20 @@
|
|||
"\n",
|
||||
" =======================> TIMING STATISTICS <=======================\n",
|
||||
"\n",
|
||||
" Total time for initialization = 4.2300E-01 seconds\n",
|
||||
" Reading cross sections = 9.3000E-02 seconds\n",
|
||||
" Total time in simulation = 1.6549E+01 seconds\n",
|
||||
" Time in transport only = 1.6535E+01 seconds\n",
|
||||
" Time in inactive batches = 2.3650E+00 seconds\n",
|
||||
" Time in active batches = 1.4184E+01 seconds\n",
|
||||
" Time synchronizing fission bank = 5.0000E-03 seconds\n",
|
||||
" Total time for initialization = 3.9900E-01 seconds\n",
|
||||
" Reading cross sections = 2.6500E-01 seconds\n",
|
||||
" Total time in simulation = 1.1488E+01 seconds\n",
|
||||
" Time in transport only = 1.1152E+01 seconds\n",
|
||||
" Time in inactive batches = 1.2180E+00 seconds\n",
|
||||
" Time in active batches = 1.0270E+01 seconds\n",
|
||||
" Time synchronizing fission bank = 4.0000E-03 seconds\n",
|
||||
" Sampling source sites = 3.0000E-03 seconds\n",
|
||||
" SEND/RECV source sites = 0.0000E+00 seconds\n",
|
||||
" SEND/RECV source sites = 1.0000E-03 seconds\n",
|
||||
" Time accumulating tallies = 0.0000E+00 seconds\n",
|
||||
" Total time for finalization = 0.0000E+00 seconds\n",
|
||||
" Total time elapsed = 1.6981E+01 seconds\n",
|
||||
" Calculation Rate (inactive) = 10570.8 neutrons/second\n",
|
||||
" Calculation Rate (active) = 7050.20 neutrons/second\n",
|
||||
" Total time for finalization = 1.0000E-03 seconds\n",
|
||||
" Total time elapsed = 1.1901E+01 seconds\n",
|
||||
" Calculation Rate (inactive) = 20525.5 neutrons/second\n",
|
||||
" Calculation Rate (active) = 9737.10 neutrons/second\n",
|
||||
"\n",
|
||||
" ============================> RESULTS <============================\n",
|
||||
"\n",
|
||||
|
|
@ -894,7 +907,7 @@
|
|||
" <td>6.250000e-07</td>\n",
|
||||
" <td>total</td>\n",
|
||||
" <td>(((total / flux) - (absorption / flux)) - (sca...</td>\n",
|
||||
" <td>-3.774758e-15</td>\n",
|
||||
" <td>-2.886580e-15</td>\n",
|
||||
" <td>0.011292</td>\n",
|
||||
" </tr>\n",
|
||||
" <tr>\n",
|
||||
|
|
@ -904,7 +917,7 @@
|
|||
" <td>2.000000e+01</td>\n",
|
||||
" <td>total</td>\n",
|
||||
" <td>(((total / flux) - (absorption / flux)) - (sca...</td>\n",
|
||||
" <td>1.443290e-15</td>\n",
|
||||
" <td>-5.551115e-16</td>\n",
|
||||
" <td>0.002570</td>\n",
|
||||
" </tr>\n",
|
||||
" </tbody>\n",
|
||||
|
|
@ -917,8 +930,8 @@
|
|||
"1 1 6.25e-07 2.00e+01 total \n",
|
||||
"\n",
|
||||
" score mean std. dev. \n",
|
||||
"0 (((total / flux) - (absorption / flux)) - (sca... -3.77e-15 1.13e-02 \n",
|
||||
"1 (((total / flux) - (absorption / flux)) - (sca... 1.44e-15 2.57e-03 "
|
||||
"0 (((total / flux) - (absorption / flux)) - (sca... -2.89e-15 1.13e-02 \n",
|
||||
"1 (((total / flux) - (absorption / flux)) - (sca... -5.55e-16 2.57e-03 "
|
||||
]
|
||||
},
|
||||
"execution_count": 22,
|
||||
|
|
@ -1167,21 +1180,21 @@
|
|||
],
|
||||
"metadata": {
|
||||
"kernelspec": {
|
||||
"display_name": "Python 2",
|
||||
"display_name": "Python 3",
|
||||
"language": "python",
|
||||
"name": "python2"
|
||||
"name": "python3"
|
||||
},
|
||||
"language_info": {
|
||||
"codemirror_mode": {
|
||||
"name": "ipython",
|
||||
"version": 2
|
||||
"version": 3
|
||||
},
|
||||
"file_extension": ".py",
|
||||
"mimetype": "text/x-python",
|
||||
"name": "python",
|
||||
"nbconvert_exporter": "python",
|
||||
"pygments_lexer": "ipython2",
|
||||
"version": "2.7.6"
|
||||
"pygments_lexer": "ipython3",
|
||||
"version": "3.5.2"
|
||||
}
|
||||
},
|
||||
"nbformat": 4,
|
||||
|
|
|
|||
File diff suppressed because one or more lines are too long
File diff suppressed because one or more lines are too long
File diff suppressed because one or more lines are too long
898
docs/source/pythonapi/examples/nuclear-data.ipynb
Normal file
898
docs/source/pythonapi/examples/nuclear-data.ipynb
Normal file
File diff suppressed because one or more lines are too long
13
docs/source/pythonapi/examples/nuclear-data.rst
Normal file
13
docs/source/pythonapi/examples/nuclear-data.rst
Normal file
|
|
@ -0,0 +1,13 @@
|
|||
.. _notebook_nuclear_data:
|
||||
|
||||
============
|
||||
Nuclear Data
|
||||
============
|
||||
|
||||
.. only:: html
|
||||
|
||||
.. notebook:: nuclear-data.ipynb
|
||||
|
||||
.. only:: latex
|
||||
|
||||
IPython notebooks must be viewed in the online HTML documentation.
|
||||
File diff suppressed because one or more lines are too long
File diff suppressed because one or more lines are too long
File diff suppressed because it is too large
Load diff
|
|
@ -27,6 +27,9 @@ Example Jupyter Notebooks
|
|||
examples/mgxs-part-ii
|
||||
examples/mgxs-part-iii
|
||||
examples/mgxs-part-iv
|
||||
examples/mdgxs-part-i
|
||||
examples/mdgxs-part-ii
|
||||
examples/nuclear-data
|
||||
|
||||
------------------------------------
|
||||
:mod:`openmc` -- Basic Functionality
|
||||
|
|
@ -35,9 +38,6 @@ Example Jupyter Notebooks
|
|||
Handling nuclear data
|
||||
---------------------
|
||||
|
||||
Classes
|
||||
+++++++
|
||||
|
||||
.. autosummary::
|
||||
:toctree: generated
|
||||
:nosignatures:
|
||||
|
|
@ -46,14 +46,6 @@ Classes
|
|||
openmc.XSdata
|
||||
openmc.MGXSLibrary
|
||||
|
||||
Functions
|
||||
+++++++++
|
||||
|
||||
.. autosummary::
|
||||
:toctree: generated
|
||||
:nosignatures:
|
||||
|
||||
openmc.ace.ascii_to_binary
|
||||
|
||||
Simulation Settings
|
||||
-------------------
|
||||
|
|
@ -224,6 +216,8 @@ Univariate Probability Distributions
|
|||
openmc.stats.Maxwell
|
||||
openmc.stats.Watt
|
||||
openmc.stats.Tabular
|
||||
openmc.stats.Legendre
|
||||
openmc.stats.Mixture
|
||||
|
||||
Angular Distributions
|
||||
---------------------
|
||||
|
|
@ -271,21 +265,41 @@ Multi-group Cross Sections
|
|||
.. autosummary::
|
||||
:toctree: generated
|
||||
:nosignatures:
|
||||
:template: myclass.rst
|
||||
:template: myclassinherit.rst
|
||||
|
||||
openmc.mgxs.MGXS
|
||||
openmc.mgxs.AbsorptionXS
|
||||
openmc.mgxs.CaptureXS
|
||||
openmc.mgxs.Chi
|
||||
openmc.mgxs.ChiPrompt
|
||||
openmc.mgxs.FissionXS
|
||||
openmc.mgxs.InverseVelocity
|
||||
openmc.mgxs.KappaFissionXS
|
||||
openmc.mgxs.MultiplicityMatrixXS
|
||||
openmc.mgxs.NuFissionXS
|
||||
openmc.mgxs.NuFissionMatrixXS
|
||||
openmc.mgxs.NuScatterXS
|
||||
openmc.mgxs.NuScatterMatrixXS
|
||||
openmc.mgxs.PromptNuFissionXS
|
||||
openmc.mgxs.ScatterXS
|
||||
openmc.mgxs.ScatterMatrixXS
|
||||
openmc.mgxs.TotalXS
|
||||
openmc.mgxs.TransportXS
|
||||
|
||||
Multi-delayed-group Cross Sections
|
||||
----------------------------------
|
||||
|
||||
.. autosummary::
|
||||
:toctree: generated
|
||||
:nosignatures:
|
||||
:template: myclassinherit.rst
|
||||
|
||||
openmc.mgxs.MDGXS
|
||||
openmc.mgxs.ChiDelayed
|
||||
openmc.mgxs.DelayedNuFissionXS
|
||||
openmc.mgxs.Beta
|
||||
openmc.mgxs.DecayRate
|
||||
|
||||
Multi-group Cross Section Libraries
|
||||
-----------------------------------
|
||||
|
||||
|
|
@ -296,6 +310,113 @@ Multi-group Cross Section Libraries
|
|||
|
||||
openmc.mgxs.Library
|
||||
|
||||
-------------------------------------
|
||||
:mod:`openmc.model` -- Model Building
|
||||
-------------------------------------
|
||||
|
||||
TRISO Fuel Modeling
|
||||
-------------------
|
||||
|
||||
Classes
|
||||
+++++++
|
||||
|
||||
.. autosummary::
|
||||
:toctree: generated
|
||||
:nosignatures:
|
||||
:template: myclass.rst
|
||||
|
||||
openmc.model.TRISO
|
||||
|
||||
Functions
|
||||
+++++++++
|
||||
|
||||
.. autosummary::
|
||||
:toctree: generated
|
||||
:nosignatures:
|
||||
|
||||
openmc.model.create_triso_lattice
|
||||
openmc.model.pack_trisos
|
||||
|
||||
--------------------------------------------
|
||||
:mod:`openmc.data` -- Nuclear Data Interface
|
||||
--------------------------------------------
|
||||
|
||||
Physical Data
|
||||
-------------
|
||||
|
||||
.. autosummary::
|
||||
:toctree: generated
|
||||
:nosignatures:
|
||||
:template: myfunction.rst
|
||||
|
||||
openmc.data.atomic_mass
|
||||
|
||||
Core Classes
|
||||
------------
|
||||
|
||||
.. autosummary::
|
||||
:toctree: generated
|
||||
:nosignatures:
|
||||
:template: myclass.rst
|
||||
|
||||
openmc.data.IncidentNeutron
|
||||
openmc.data.Reaction
|
||||
openmc.data.Product
|
||||
openmc.data.Tabulated1D
|
||||
openmc.data.ThermalScattering
|
||||
openmc.data.CoherentElastic
|
||||
openmc.data.FissionEnergyRelease
|
||||
openmc.data.DataLibrary
|
||||
|
||||
Angle-Energy Distributions
|
||||
--------------------------
|
||||
|
||||
.. autosummary::
|
||||
:toctree: generated
|
||||
:nosignatures:
|
||||
:template: myclass.rst
|
||||
|
||||
openmc.data.AngleEnergy
|
||||
openmc.data.KalbachMann
|
||||
openmc.data.CorrelatedAngleEnergy
|
||||
openmc.data.UncorrelatedAngleEnergy
|
||||
openmc.data.NBodyPhaseSpace
|
||||
openmc.data.AngleDistribution
|
||||
openmc.data.EnergyDistribution
|
||||
openmc.data.ArbitraryTabulated
|
||||
openmc.data.GeneralEvaporation
|
||||
openmc.data.MaxwellEnergy
|
||||
openmc.data.Evaporation
|
||||
openmc.data.WattEnergy
|
||||
openmc.data.MadlandNix
|
||||
openmc.data.DiscretePhoton
|
||||
openmc.data.LevelInelastic
|
||||
openmc.data.ContinuousTabular
|
||||
|
||||
ACE Format
|
||||
----------
|
||||
|
||||
Classes
|
||||
+++++++
|
||||
|
||||
.. autosummary::
|
||||
:toctree: generated
|
||||
:nosignatures:
|
||||
:template: myclass.rst
|
||||
|
||||
openmc.data.ace.Library
|
||||
openmc.data.ace.Table
|
||||
|
||||
Functions
|
||||
+++++++++
|
||||
|
||||
.. autosummary::
|
||||
:toctree: generated
|
||||
:nosignatures:
|
||||
|
||||
openmc.data.ace.ascii_to_binary
|
||||
openmc.data.write_compact_458_library
|
||||
|
||||
.. _Jupyter: https://jupyter.org/
|
||||
.. _NumPy: http://www.numpy.org/
|
||||
.. _Codecademy: https://www.codecademy.com/tracks/python
|
||||
|
|
|
|||
|
|
@ -1,78 +1,83 @@
|
|||
.. _releasenotes:
|
||||
|
||||
==============================
|
||||
Release Notes for OpenMC 0.7.1
|
||||
Release Notes for OpenMC 0.8.0
|
||||
==============================
|
||||
|
||||
This release of OpenMC provides some substantial improvements over version
|
||||
0.7.0. Non-simple cell regions can now be defined through the ``|`` (union) and
|
||||
``~`` (complement) operators. Similar changes in the Python API also allow
|
||||
complex cell regions to be defined. A true secondary particle bank now exists;
|
||||
this is crucial for photon transport (to be added in the next minor release). A
|
||||
rich API for multi-group cross section generation has been added via the
|
||||
``openmc.mgxs`` Python module.
|
||||
This release of OpenMC includes a few new major features including the
|
||||
capability to perform neutron transport with multi-group cross section data as
|
||||
well as experimental support for the windowed multipole method being developed
|
||||
at MIT. Source sampling options have also been expanded significantly, with the
|
||||
option to supply arbitrary tabular and discrete distributions for energy, angle,
|
||||
and spatial coordinates.
|
||||
|
||||
Various improvements to tallies have also been made. It is now possible to
|
||||
explicitly specify that a collision estimator be used in a tally. A new
|
||||
``delayedgroup`` filter and ``delayed-nu-fission`` score allow a user to obtain
|
||||
delayed fission neutron production rates filtered by delayed group. Finally, the
|
||||
new ``inverse-velocity`` score may be useful for calculating kinetics
|
||||
parameters.
|
||||
The Python API has been significantly restructured in this release compared to
|
||||
version 0.7.1. Any scripts written based on the version 0.7.1 API will likely
|
||||
need to be rewritten. Some of the most visible changes include the following:
|
||||
|
||||
.. caution:: In previous versions, depending on how OpenMC was compiled binary
|
||||
output was either given in HDF5 or a flat binary format. With this
|
||||
version, all binary output is now HDF5 which means you **must**
|
||||
have HDF5 in order to install OpenMC. Please consult the user's
|
||||
guide for instructions on how to compile with HDF5.
|
||||
- ``SettingsFile`` is now ``Settings``, ``MaterialsFile`` is now ``Materials``,
|
||||
and ``TalliesFile`` is now ``Tallies``.
|
||||
- The ``GeometryFile`` class no longer exists and is replaced by the
|
||||
``Geometry`` class which now has an ``export_to_xml()`` method.
|
||||
- Source distributions are defined using the ``Source`` class and assigned to
|
||||
the ``Settings.source`` property.
|
||||
- The ``Executor`` class no longer exists and is replaced by ``openmc.run()``
|
||||
and ``openmc.plot_geometry()`` functions.
|
||||
|
||||
The Python API documentation has also been significantly expanded.
|
||||
|
||||
-------------------
|
||||
System Requirements
|
||||
-------------------
|
||||
|
||||
There are no special requirements for running the OpenMC code. As of this
|
||||
release, OpenMC has been tested on a variety of Linux distributions, Mac OS X,
|
||||
and Microsoft Windows 7. Memory requirements will vary depending on the size of
|
||||
the problem at hand (mostly on the number of nuclides in the problem).
|
||||
release, OpenMC has been tested on a variety of Linux distributions and Mac
|
||||
OS X. Numerous users have reported working builds on Microsoft Windows, but your
|
||||
mileage may vary. Memory requirements will vary depending on the size of the
|
||||
problem at hand (mostly on the number of nuclides and tallies in the problem).
|
||||
|
||||
------------
|
||||
New Features
|
||||
------------
|
||||
|
||||
- Support for complex cell regions (union and complement operators)
|
||||
- Generic quadric surface type
|
||||
- Improved handling of secondary particles
|
||||
- Binary output is now solely HDF5
|
||||
- ``openmc.mgxs`` Python module enabling multi-group cross section generation
|
||||
- Collision estimator for tallies
|
||||
- Delayed fission neutron production tallies with ability to filter by delayed
|
||||
group
|
||||
- Inverse velocity tally score
|
||||
- Performance improvements for binary search
|
||||
- Performance improvements for reaction rate tallies
|
||||
- Multi-group mode
|
||||
- Vast improvements to the Python API
|
||||
- Experimental windowed multipole capability
|
||||
- Periodic boundary conditions
|
||||
- Expanded source sampling options
|
||||
- Distributed materials
|
||||
- Subcritical multiplication support
|
||||
- Improved method for reproducible URR table sampling
|
||||
- Refactor of continuous-energy reaction data
|
||||
- Improved documentation and new Jupyter notebooks
|
||||
|
||||
---------
|
||||
Bug Fixes
|
||||
---------
|
||||
|
||||
- 299322_: Bug with material filter when void material present
|
||||
- d74840_: Fix triggers on tallies with multiple filters
|
||||
- c29a81_: Correctly handle maximum transport energy
|
||||
- 3edc23_: Fixes in the nu-scatter score
|
||||
- 629e3b_: Assume unspecified surface coefficients are zero in Python API
|
||||
- 5dbe8b_: Fix energy filters for openmc-plot-mesh-tally
|
||||
- ff66f4_: Fixes in the openmc-plot-mesh-tally script
|
||||
- 441fd4_: Fix bug in kappa-fission score
|
||||
- 7e5974_: Allow fixed source simulations from Python API
|
||||
- 70daa7_: Make sure MT=3 cross section is not used
|
||||
- 40b05f_: Ensure source bank is resampled for fixed source runs
|
||||
- 9586ed_: Fix two hexagonal lattice bugs
|
||||
- a855e8_: Make sure graphite models don't error out on max events
|
||||
- 7294a1_: Fix incorrect check on cmfd.xml
|
||||
- 12f246_: Ensure number of realizations is written to statepoint
|
||||
- 0227f4_: Fix bug when sampling multiple energy distributions
|
||||
- 51deaa_: Prevent segfault when user specifies '18' on tally scores
|
||||
- fed74b_: Prevent duplicate tally scores
|
||||
- 8467ae_: Better threshold for allowable lost particles
|
||||
- 493c6f_: Fix type of return argument for h5pget_driver_f
|
||||
|
||||
.. _299322: https://github.com/mit-crpg/openmc/commit/299322
|
||||
.. _d74840: https://github.com/mit-crpg/openmc/commit/d74840
|
||||
.. _c29a81: https://github.com/mit-crpg/openmc/commit/c29a81
|
||||
.. _3edc23: https://github.com/mit-crpg/openmc/commit/3edc23
|
||||
.. _629e3b: https://github.com/mit-crpg/openmc/commit/629e3b
|
||||
.. _5dbe8b: https://github.com/mit-crpg/openmc/commit/5dbe8b
|
||||
.. _ff66f4: https://github.com/mit-crpg/openmc/commit/ff66f4
|
||||
.. _441fd4: https://github.com/mit-crpg/openmc/commit/441fd4
|
||||
.. _7e5974: https://github.com/mit-crpg/openmc/commit/7e5974
|
||||
.. _70daa7: https://github.com/mit-crpg/openmc/commit/70daa7
|
||||
.. _40b05f: https://github.com/mit-crpg/openmc/commit/40b05f
|
||||
.. _9586ed: https://github.com/mit-crpg/openmc/commit/9586ed
|
||||
.. _a855e8: https://github.com/mit-crpg/openmc/commit/a855e8
|
||||
.. _7294a1: https://github.com/mit-crpg/openmc/commit/7294a1
|
||||
.. _12f246: https://github.com/mit-crpg/openmc/commit/12f246
|
||||
.. _0227f4: https://github.com/mit-crpg/openmc/commit/0227f4
|
||||
.. _51deaa: https://github.com/mit-crpg/openmc/commit/51deaa
|
||||
.. _fed74b: https://github.com/mit-crpg/openmc/commit/fed74b
|
||||
.. _8467ae: https://github.com/mit-crpg/openmc/commit/8467ae
|
||||
.. _493c6f: https://github.com/mit-crpg/openmc/commit/493c6f
|
||||
|
||||
------------
|
||||
Contributors
|
||||
|
|
@ -81,11 +86,11 @@ Contributors
|
|||
This release contains new contributions from the following people:
|
||||
|
||||
- `Will Boyd <wbinventor@gmail.com>`_
|
||||
- `Sterling Harper <sterlingmharper@mit.edu>`_
|
||||
- `Bryan Herman <hermab53@gmail.com>`_
|
||||
- `Derek Gaston <friedmud@gmail.com>`_
|
||||
- `Sterling Harper <sterlingmharper@gmail.com>`_
|
||||
- `Colin Josey <cjosey@mit.edu>`_
|
||||
- `Jingang Liang <liangjg2008@gmail.com>`_
|
||||
- `Adam Nelson <nelsonag@umich.edu>`_
|
||||
- `Paul Romano <paul.k.romano@gmail.com>`_
|
||||
- `Kelly Rowland <kellylynnerowland@gmail.com>`_
|
||||
- `Sam Shaner <samuelshaner@gmail.com>`_
|
||||
- `Jon Walsh <walshjon@mit.edu>`_
|
||||
|
|
|
|||
|
|
@ -65,7 +65,7 @@ Now let's look at the pros and cons of Monte Carlo methods:
|
|||
|
||||
- **Pro**: Running simulations in parallel is conceptually very simple.
|
||||
|
||||
- **Con**: Because they related on repeated random sampling, they are
|
||||
- **Con**: Because they rely on repeated random sampling, they are
|
||||
computationally very expensive.
|
||||
|
||||
- **Con**: A simulation doesn't automatically give you the global solution
|
||||
|
|
|
|||
|
|
@ -281,6 +281,8 @@ based on the recommended value in LA-UR-14-24530_.
|
|||
|
||||
.. note:: This element is not used in the multi-group :ref:`energy_mode`.
|
||||
|
||||
.. _multipole_library:
|
||||
|
||||
``<multipole_library>`` Element
|
||||
-------------------------------
|
||||
|
||||
|
|
@ -290,8 +292,8 @@ OpenMC can use it for on-the-fly Doppler-broadening of resolved resonance range
|
|||
cross sections. If this element is absent from the settings.xml file, the
|
||||
:envvar:`OPENMC_MULTIPOLE_LIBRARY` environment variable will be used.
|
||||
|
||||
.. note:: The <use_windowed_multipole> element must also be set to "true"
|
||||
for windowed multipole functionality.
|
||||
.. note:: The :ref:`temperature_method` must also be set to "multipole" for
|
||||
windowed multipole functionality.
|
||||
|
||||
``<max_order>`` Element
|
||||
---------------------------
|
||||
|
|
@ -395,19 +397,16 @@ attributes or sub-elements:
|
|||
|
||||
:scatterer:
|
||||
An element with attributes/sub-elements called ``nuclide``, ``method``,
|
||||
``xs_label``, ``xs_label_0K``, ``E_min``, and ``E_max``. The ``nuclide``
|
||||
attribute is the name, as given by the ``name`` attribute within the
|
||||
``nuclide`` sub-element of the ``material`` element in ``materials.xml``,
|
||||
of the nuclide to which a resonance scattering treatment is to be applied.
|
||||
``E_min``, and ``E_max``. The ``nuclide`` attribute is the name, as given
|
||||
by the ``name`` attribute within the ``nuclide`` sub-element of the
|
||||
``material`` element in ``materials.xml``, of the nuclide to which a
|
||||
resonance scattering treatment is to be applied.
|
||||
The ``method`` attribute gives the type of resonance scattering treatment
|
||||
that is to be applied to the ``nuclide``. Acceptable inputs - none of
|
||||
which are case-sensitive - for the ``method`` attribute are ``ARES``,
|
||||
``CXS``, ``WCM``, and ``DBRC``. Descriptions of each of these methods
|
||||
are documented here_. The ``xs_label`` attribute gives the label for the
|
||||
cross section data of the ``nuclide`` at a given temperature. The
|
||||
``xs_label_0K`` gives the label for the 0 K cross section data for the
|
||||
``nuclide``. The ``E_min`` attribute gives the minimum energy above
|
||||
which the ``method`` is applied. The ``E_max`` attribute gives the
|
||||
are documented here_. The ``E_min`` attribute gives the minimum energy
|
||||
above which the ``method`` is applied. The ``E_max`` attribute gives the
|
||||
maximum energy below which the ``method`` is applied. One example would
|
||||
be as follows:
|
||||
|
||||
|
|
@ -419,16 +418,12 @@ attributes or sub-elements:
|
|||
<scatterer>
|
||||
<nuclide>U-238</nuclide>
|
||||
<method>ARES</method>
|
||||
<xs_label>92238.72c</xs_label>
|
||||
<xs_label_0K>92238.00c</xs_label_0K>
|
||||
<E_min>5.0e-6</E_min>
|
||||
<E_max>40.0e-6</E_max>
|
||||
</scatterer>
|
||||
<scatterer>
|
||||
<nuclide>Pu-239</nuclide>
|
||||
<method>dbrc</method>
|
||||
<xs_label>94239.72c</xs_label>
|
||||
<xs_label_0K>94239.00c</xs_label_0K>
|
||||
<E_min>0.01e-6</E_min>
|
||||
<E_max>210.0e-6</E_max>
|
||||
</scatterer>
|
||||
|
|
@ -688,7 +683,7 @@ attributes/sub-elements:
|
|||
|
||||
*Default*: false
|
||||
|
||||
:source_write:
|
||||
: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
|
||||
size of state points if large numbers of particles per batch are used.
|
||||
|
|
@ -714,6 +709,48 @@ survival biasing, otherwise known as implicit capture or absorption.
|
|||
|
||||
*Default*: false
|
||||
|
||||
.. _temperature_default:
|
||||
|
||||
``<temperature_default>`` Element
|
||||
---------------------------------
|
||||
|
||||
The ``<temperature_default>`` element specifies a default temperature in Kelvin
|
||||
that is to be applied to cells in the absence of an explicit cell temperature or
|
||||
a material default temperature.
|
||||
|
||||
*Default*: 293.6 K
|
||||
|
||||
.. _temperature_method:
|
||||
|
||||
``<temperature_method>`` Element
|
||||
--------------------------------
|
||||
|
||||
The ``<temperature_method>`` element has an accepted value of "nearest",
|
||||
"interpolation", or "multipole". A value of "nearest" indicates that for each
|
||||
cell, the nearest temperature at which cross sections are given is to be
|
||||
applied, within a given tolerance (see :ref:`temperature_tolerance`). A value of
|
||||
"interpolation" indicates that cross sections are to be linear-linear
|
||||
interpolated between temperatures at which nuclear data are present (see
|
||||
:ref:`temperature_treatment`). A value of "multipole" indicates that the
|
||||
windowed multipole method should be used to evaluate temperature-dependent cross
|
||||
sections in the resolved resonance range (a :ref:`windowed multipole library
|
||||
<multipole_library>` must also be available).
|
||||
|
||||
*Default*: "nearest"
|
||||
|
||||
.. _temperature_tolerance:
|
||||
|
||||
``<temperature_tolerance>`` Element
|
||||
-----------------------------------
|
||||
|
||||
The ``<temperature_tolerance>`` element specifies a tolerance in Kelvin that is
|
||||
to be applied when the "nearest" temperature method is used. For example, if a
|
||||
cell temperature is 340 K and the tolerance is 15 K, then the closest
|
||||
temperature in the range of 325 K to 355 K will be used to evaluate cross
|
||||
sections.
|
||||
|
||||
*Default*: 10 K
|
||||
|
||||
``<threads>`` Element
|
||||
---------------------
|
||||
|
||||
|
|
@ -836,6 +873,35 @@ displayed. This element takes the following attributes:
|
|||
|
||||
*Default*: 5
|
||||
|
||||
``<volume_calc>`` Element
|
||||
-------------------------
|
||||
|
||||
The ``<volume_calc>`` element indicates that a stochastic volume calculation
|
||||
should be run at the beginning of the simulation. This element has the following
|
||||
sub-elements/attributes:
|
||||
|
||||
:cells:
|
||||
The unique IDs of cells for which the volume should be estimated.
|
||||
|
||||
*Default*: None
|
||||
|
||||
:samples:
|
||||
The number of samples used to estimate volumes.
|
||||
|
||||
*Default*: None
|
||||
|
||||
:lower_left:
|
||||
The lower-left Cartesian coordinates of a bounding box that is used to
|
||||
sample points within.
|
||||
|
||||
*Default*: None
|
||||
|
||||
:upper_right:
|
||||
The upper-right Cartesian coordinates of a bounding box that is used to
|
||||
sample points within.
|
||||
|
||||
*Default*: None
|
||||
|
||||
--------------------------------------
|
||||
Geometry Specification -- geometry.xml
|
||||
--------------------------------------
|
||||
|
|
@ -921,11 +987,19 @@ Each ``<surface>`` element can have the following attributes or sub-elements:
|
|||
*Default*: None
|
||||
|
||||
:boundary:
|
||||
The boundary condition for the surface. This can be "transmission",
|
||||
"vacuum", or "reflective".
|
||||
The boundary condition for the surface. This can be "transmission",
|
||||
"vacuum", "reflective", or "periodic". Periodic boundary conditions can
|
||||
only be applied to x-, y-, and z-planes. Only axis-aligned periodicity is
|
||||
supported, i.e., x-planes can only be paired with x-planes. Specify which
|
||||
planes are periodic and the code will automatically identify which planes
|
||||
are paired together.
|
||||
|
||||
*Default*: "transmission"
|
||||
|
||||
:periodic_surface_id:
|
||||
If a periodic boundary condition is applied, this attribute identifies the
|
||||
``id`` of the corresponding periodic sufrace.
|
||||
|
||||
The following quadratic surfaces can be modeled:
|
||||
|
||||
:x-plane:
|
||||
|
|
@ -1053,7 +1127,9 @@ Each ``<cell>`` element can have the following attributes or sub-elements:
|
|||
specified for the "distributed temperature" feature. This will give each
|
||||
unique instance of the cell its own temperature.
|
||||
|
||||
*Default*: The temperature of the coldest nuclide in the cell's material(s)
|
||||
*Default*: If a material default temperature is supplied, it is used. In the
|
||||
absence of a material default temperature, the :ref:`global default
|
||||
temperature <temperature_default>` is used.
|
||||
|
||||
:rotation:
|
||||
If the cell is filled with a universe, this element specifies the angles in
|
||||
|
|
@ -1258,6 +1334,14 @@ Each ``material`` element can have the following attributes or sub-elements:
|
|||
|
||||
*Default*: ""
|
||||
|
||||
:temperature:
|
||||
An element with no attributes which is used to set the default temperature
|
||||
of the material in Kelvin.
|
||||
|
||||
*Default*: If a material default temperature is not given and a cell
|
||||
temperature is not specified, the :ref:`global default temperature
|
||||
<temperature_default>` is used.
|
||||
|
||||
:density:
|
||||
An element with attributes/sub-elements called ``value`` and ``units``. The
|
||||
``value`` attribute is the numeric value of the density while the ``units``
|
||||
|
|
@ -1278,17 +1362,16 @@ Each ``material`` element can have the following attributes or sub-elements:
|
|||
``nuclide``, ``element``, or ``sab`` quantity.
|
||||
|
||||
:nuclide:
|
||||
An element with attributes/sub-elements called ``name``, ``xs``, and ``ao``
|
||||
An element with attributes/sub-elements called ``name``, and ``ao``
|
||||
or ``wo``. The ``name`` attribute is the name of the cross-section for a
|
||||
desired nuclide while the ``xs`` attribute is the cross-section
|
||||
identifier. Finally, the ``ao`` and ``wo`` attributes specify the atom or
|
||||
desired nuclide. Finally, the ``ao`` and ``wo`` attributes specify the atom or
|
||||
weight percent of that nuclide within the material, respectively. One
|
||||
example would be as follows:
|
||||
|
||||
.. code-block:: xml
|
||||
|
||||
<nuclide name="H-1" xs="70c" ao="2.0" />
|
||||
<nuclide name="O-16" xs="70c" ao="1.0" />
|
||||
<nuclide name="H1" ao="2.0" />
|
||||
<nuclide name="O16" ao="1.0" />
|
||||
|
||||
.. note:: If one nuclide is specified in atom percent, all others must also
|
||||
be given in atom percent. The same applies for weight percentages.
|
||||
|
|
@ -1312,11 +1395,10 @@ Each ``material`` element can have the following attributes or sub-elements:
|
|||
Specifies that a natural element is present in the material. The natural
|
||||
element is split up into individual isotopes based on `IUPAC Isotopic
|
||||
Compositions of the Elements 2009`_. This element has
|
||||
attributes/sub-elements called ``name``, ``xs``, and ``ao``. The ``name``
|
||||
attribute is the atomic symbol of the element while the ``xs`` attribute is
|
||||
the cross-section identifier. Finally, the ``ao`` attribute specifies the
|
||||
atom percent of the element within the material, respectively. One example
|
||||
would be as follows:
|
||||
attributes/sub-elements called ``name``, and ``ao``. The ``name``
|
||||
attribute is the atomic symbol of the element. Finally, the ``ao``
|
||||
attribute specifies the atom percent of the element within the material,
|
||||
respectively. One example would be as follows:
|
||||
|
||||
.. code-block:: xml
|
||||
|
||||
|
|
@ -1346,10 +1428,9 @@ Each ``material`` element can have the following attributes or sub-elements:
|
|||
multi-group :ref:`energy_mode`.
|
||||
|
||||
:sab:
|
||||
Associates an S(a,b) table with the material. This element has
|
||||
attributes/sub-elements called ``name`` and ``xs``. The ``name`` attribute
|
||||
is the name of the S(a,b) table that should be associated with the material,
|
||||
and ``xs`` is the cross-section identifier for the table.
|
||||
Associates an S(a,b) table with the material. This element has one
|
||||
attribute/sub-element called ``name``. The ``name`` attribute
|
||||
is the name of the S(a,b) table that should be associated with the material.
|
||||
|
||||
*Default*: None
|
||||
|
||||
|
|
@ -1360,14 +1441,13 @@ Each ``material`` element can have the following attributes or sub-elements:
|
|||
recognizes that some multi-group libraries may be providing material
|
||||
specific macroscopic cross sections instead of always providing nuclide
|
||||
specific data like in the continuous-energy case. To that end, the
|
||||
macroscopic element has attributes/sub-elements called ``name``, and ``xs``.
|
||||
macroscopic element has one attribute/sub-element called ``name``.
|
||||
The ``name`` attribute is the name of the cross-section for a
|
||||
desired nuclide while the ``xs`` attribute is the cross-section
|
||||
identifier. One example would be as follows:
|
||||
desired nuclide. One example would be as follows:
|
||||
|
||||
.. code-block:: xml
|
||||
|
||||
<macroscopic name="UO2" xs="71c" />
|
||||
<macroscopic name="UO2" />
|
||||
|
||||
.. note:: This element is only used in the multi-group :ref:`energy_mode`.
|
||||
|
||||
|
|
@ -1376,18 +1456,6 @@ Each ``material`` element can have the following attributes or sub-elements:
|
|||
.. _IUPAC Isotopic Compositions of the Elements 2009:
|
||||
http://pac.iupac.org/publications/pac/pdf/2011/pdf/8302x0397.pdf
|
||||
|
||||
``<default_xs>`` Element
|
||||
------------------------
|
||||
|
||||
In some circumstances, the cross-section identifier may be the same for many or
|
||||
all nuclides in a given problem. In this case, rather than specifying the
|
||||
``xs=...`` attribute on every nuclide, a ``<default_xs>`` element can be used to
|
||||
set the default cross-section identifier for any nuclide without an identifier
|
||||
explicitly listed. This element has no attributes and accepts a 3-letter string
|
||||
that indicates the default cross-section identifier, e.g. "70c".
|
||||
|
||||
*Default*: None
|
||||
|
||||
------------------------------------
|
||||
Tallies Specification -- tallies.xml
|
||||
------------------------------------
|
||||
|
|
@ -1623,7 +1691,8 @@ The ``<tally>`` element accepts the following sub-elements:
|
|||
|Score | Description |
|
||||
+======================+===================================================+
|
||||
|absorption |Total absorption rate. This accounts for all |
|
||||
| |reactions which do not produce secondary neutrons. |
|
||||
| |reactions which do not produce secondary neutrons |
|
||||
| |as well as fission. |
|
||||
+----------------------+---------------------------------------------------+
|
||||
|elastic |Elastic scattering reaction rate. |
|
||||
+----------------------+---------------------------------------------------+
|
||||
|
|
@ -1755,6 +1824,10 @@ The ``<tally>`` element accepts the following sub-elements:
|
|||
| |fission. This score type is not used in the |
|
||||
| |multi-group :ref:`energy_mode`. |
|
||||
+----------------------+---------------------------------------------------+
|
||||
|prompt-nu-fission |Total production of prompt neutrons due to |
|
||||
| |fission. This score type is not used in the |
|
||||
| |multi-group :ref:`energy_mode`. |
|
||||
+----------------------+---------------------------------------------------+
|
||||
|nu-fission |Total production of neutrons due to fission. |
|
||||
+----------------------+---------------------------------------------------+
|
||||
|nu-scatter, |These scores are similar in functionality to their |
|
||||
|
|
@ -1796,6 +1869,32 @@ The ``<tally>`` element accepts the following sub-elements:
|
|||
| |:math:`\gamma`-rays are assumed to deposit their |
|
||||
| |energy locally. Units are MeV per source particle. |
|
||||
+----------------------+---------------------------------------------------+
|
||||
|fission-q-prompt |The prompt fission energy production rate. This |
|
||||
| |energy comes in the form of fission fragment |
|
||||
| |nuclei, prompt neutrons, and prompt |
|
||||
| |:math:`\gamma`-rays. This value depends on the |
|
||||
| |incident energy and it requires that the nuclear |
|
||||
| |data library contains the optional fission energy |
|
||||
| |release data. Energy is assumed to be deposited |
|
||||
| |locally. Units are MeV per source particle. |
|
||||
+----------------------+---------------------------------------------------+
|
||||
|fission-q-recoverable |The recoverable fission energy production rate. |
|
||||
| |This energy comes in the form of fission fragment |
|
||||
| |nuclei, prompt and delayed neutrons, prompt and |
|
||||
| |delayed :math:`\gamma`-rays, and delayed |
|
||||
| |:math:`\beta`-rays. This tally differs from the |
|
||||
| |kappa-fission tally in that it is dependent on |
|
||||
| |incident neutron energy and it requires that the |
|
||||
| |nuclear data library contains the optional fission |
|
||||
| |energy release data. Energy is assumed to be |
|
||||
| |deposited locally. Units are MeV per source |
|
||||
| |paticle. |
|
||||
+----------------------+---------------------------------------------------+
|
||||
|decay-rate |The delayed-nu-fission-weighted decay rate where |
|
||||
| |the decay rate is in units of inverse seconds. |
|
||||
| |This score type is not used in the |
|
||||
| |multi-group :ref:`energy_mode`. |
|
||||
+----------------------+---------------------------------------------------+
|
||||
|
||||
.. note::
|
||||
The ``analog`` estimator is actually identical to the ``collision``
|
||||
|
|
@ -2077,8 +2176,8 @@ attributes or sub-elements. These are not used in "voxel" plots:
|
|||
*Default*: None
|
||||
|
||||
:meshlines:
|
||||
The ``meshlines`` sub-element allows for plotting the boundaries of
|
||||
a tally mesh on top of a plot. Only one ``meshlines`` element is allowed per
|
||||
The ``meshlines`` sub-element allows for plotting the boundaries of a
|
||||
regular mesh on top of a plot. Only one ``meshlines`` element is allowed per
|
||||
``plot`` element, and it must contain as attributes or sub-elements a mesh
|
||||
type and a linewidth. Optionally, a color may be specified for the overlay:
|
||||
|
||||
|
|
|
|||
|
|
@ -204,20 +204,22 @@ should be used:
|
|||
Compiling with MPI
|
||||
++++++++++++++++++
|
||||
|
||||
To compile with MPI, set the :envvar:`FC` environment variable to the path to
|
||||
the MPI Fortran wrapper. For example, in a bash shell:
|
||||
To compile with MPI, set the :envvar:`FC` and :envvar:`CC` environment variables
|
||||
to the path to the MPI Fortran and C wrappers, respectively. For example, in a
|
||||
bash shell:
|
||||
|
||||
.. code-block:: sh
|
||||
|
||||
export FC=mpif90
|
||||
export CC=mpicc
|
||||
cmake /path/to/openmc
|
||||
|
||||
Note that in many shells, an environment variable can be set for a single
|
||||
command, i.e.
|
||||
Note that in many shells, environment variables can be set for a single command,
|
||||
i.e.
|
||||
|
||||
.. code-block:: sh
|
||||
|
||||
FC=mpif90 cmake /path/to/openmc
|
||||
FC=mpif90 CC=mpicc cmake /path/to/openmc
|
||||
|
||||
Selecting HDF5 Installation
|
||||
+++++++++++++++++++++++++++
|
||||
|
|
@ -343,7 +345,7 @@ compiler, it is necessary to specify that all objects be compiled with the
|
|||
.. code-block:: sh
|
||||
|
||||
mkdir build && cd build
|
||||
FC=ifort FFLAGS=-mmic cmake -Dopenmp=on ..
|
||||
FC=ifort CC=icc FFLAGS=-mmic cmake -Dopenmp=on ..
|
||||
make
|
||||
|
||||
Note that unless an HDF5 build for the Intel Xeon Phi is already on your target
|
||||
|
|
@ -381,14 +383,16 @@ Cross Section Configuration
|
|||
---------------------------
|
||||
|
||||
In order to run a simulation with OpenMC, you will need cross section data for
|
||||
each nuclide or material in your problem. OpenMC can be run in
|
||||
continuous-energy or multi-group mode.
|
||||
each nuclide or material in your problem. OpenMC can be run in continuous-energy
|
||||
or multi-group mode.
|
||||
|
||||
In continuous-energy mode OpenMC uses ACE format cross sections; in this case
|
||||
you can use nuclear data that was processed with NJOY_, such as that
|
||||
distributed with MCNP_ or Serpent_. Several sources provide free processed
|
||||
ACE data as described below. The TALYS-based evaluated nuclear data library,
|
||||
TENDL_, is also openly available in ACE format.
|
||||
In continuous-energy mode, OpenMC uses a native HDF5 format to store all nuclear
|
||||
data. If you have ACE format data that was produced with NJOY_, such as that
|
||||
distributed with MCNP_ or Serpent_, it can be converted to the HDF5 format using
|
||||
the :ref:`openmc-ace-to-hdf5 <other_cross_sections>` script distributed with
|
||||
OpenMC. Several sources provide openly available ACE data as described
|
||||
below. The TALYS-based evaluated nuclear data library, TENDL_, is also available
|
||||
in ACE format.
|
||||
|
||||
In multi-group mode, OpenMC utilizes an XML-based library format which can be
|
||||
used to describe nuclide- or material-specific quantities.
|
||||
|
|
@ -398,8 +402,8 @@ Using ENDF/B-VII.1 Cross Sections from NNDC
|
|||
|
||||
The NNDC_ provides ACE data from the ENDF/B-VII.1 neutron and thermal scattering
|
||||
sublibraries at four temperatures processed using NJOY_. To use this data with
|
||||
OpenMC, a script is provided with OpenMC that will automatically download,
|
||||
extract, and set up a confiuration file:
|
||||
OpenMC, a script is provided with OpenMC that will automatically download and
|
||||
extract the ACE data, fix any deficiencies, and create an HDF5 library:
|
||||
|
||||
.. code-block:: sh
|
||||
|
||||
|
|
@ -408,56 +412,99 @@ extract, and set up a confiuration file:
|
|||
|
||||
At this point, you should set the :envvar:`OPENMC_CROSS_SECTIONS` environment
|
||||
variable to the absolute path of the file
|
||||
``openmc/data/nndc/cross_sections.xml``. This cross section set is used by the
|
||||
test suite.
|
||||
``openmc/data/nndc_hdf5/cross_sections.xml``. This cross section set is used by
|
||||
the test suite.
|
||||
|
||||
Using JEFF Cross Sections from OECD/NEA
|
||||
---------------------------------------
|
||||
|
||||
The NEA_ provides processed ACE data from the JEFF_ nuclear library upon
|
||||
request. A DVD of the data can be requested here_. To use this data with OpenMC,
|
||||
the following steps must be taken:
|
||||
The NEA_ provides processed ACE data from the JEFF_ library. To use this data
|
||||
with OpenMC, a script is provided with OpenMC that will automatically download
|
||||
and extract the ACE data, fix any deficiencies, and create an HDF5 library.
|
||||
|
||||
1. Copy and unzip the data on the DVD to a directory on your computer.
|
||||
2. In the root directory, a file named ``xsdir``, or some variant thereof,
|
||||
should be present. This file contains a listing of all the cross sections and
|
||||
is used by MCNP. This file should be converted to a ``cross_sections.xml``
|
||||
file for use with OpenMC. A utility is provided in the OpenMC distribution
|
||||
for this purpose:
|
||||
.. code-block:: sh
|
||||
|
||||
.. code-block:: sh
|
||||
cd openmc/data
|
||||
python get_jeff_data.py
|
||||
|
||||
openmc/scripts/openmc-xsdir-to-xml xsdir31 cross_sections.xml
|
||||
|
||||
3. In the converted ``cross_sections.xml`` file, change the contents of the
|
||||
<directory> element to the absolute path of the directory containing the
|
||||
actual ACE files.
|
||||
4. Additionally, you may need to change any occurrences of upper-case "ACE"
|
||||
within the ``cross_sections.xml`` file to lower-case.
|
||||
5. Either set the :ref:`cross_sections` in a settings.xml file or the
|
||||
:envvar:`OPENMC_CROSS_SECTIONS` environment variable to the absolute path of
|
||||
the ``cross_sections.xml`` file.
|
||||
At this point, you should set the :envvar:`OPENMC_CROSS_SECTIONS` environment
|
||||
variable to the absolute path of the file
|
||||
``openmc/data/jeff-3.2-hdf5/cross_sections.xml``.
|
||||
|
||||
Using Cross Sections from MCNP
|
||||
------------------------------
|
||||
|
||||
To use cross sections distributed with MCNP, change the <directory> element in
|
||||
the ``cross_sections.xml`` file in the root directory of the OpenMC distribution
|
||||
to the location of the MCNP cross sections. Then, either set the
|
||||
:ref:`cross_sections` in a settings.xml file or the
|
||||
:envvar:`OPENMC_CROSS_SECTIONS` environment variable to the absolute path of
|
||||
the ``cross_sections.xml`` file.
|
||||
OpenMC is provided with a script that will automatically convert ENDF/B-VII.0
|
||||
and ENDF/B-VII.1 ACE data that is provided with MCNP5 or MCNP6. To convert the
|
||||
ENDF/B-VII.0 ACE files (``endf70[a-k]`` and ``endf70sab``) into the native HDF5
|
||||
format, run the following:
|
||||
|
||||
Using Cross Sections from Serpent
|
||||
---------------------------------
|
||||
.. code-block:: sh
|
||||
|
||||
cd openmc/data
|
||||
python convert_mcnp_endf70.py /path/to/mcnpdata/
|
||||
|
||||
where ``/path/to/mcnpdata`` is the directory containing the ``endf70[a-k]``
|
||||
files.
|
||||
|
||||
To convert the ENDF/B-VII.1 ACE files (the endf71x and ENDF71SaB libraries), use
|
||||
the following script:
|
||||
|
||||
.. code-block:: sh
|
||||
|
||||
cd openmc/data
|
||||
python convert_mcnp_endf71.py /path/to/mcnpdata
|
||||
|
||||
where ``/path/to/mcnpdata`` is the directory containing the ``endf71x`` and
|
||||
``ENDF71SaB`` directories.
|
||||
|
||||
.. _other_cross_sections:
|
||||
|
||||
Using Other Cross Sections
|
||||
--------------------------
|
||||
|
||||
If you have a library of ACE format cross sections other than those listed above
|
||||
that you need to convert to OpenMC's HDF5 format, the ``openmc-ace-to-hdf5``
|
||||
script can be used. There are four different ways you can specify ACE libraries
|
||||
that are to be converted:
|
||||
|
||||
1. List each ACE library as a positional argument. This is very useful in
|
||||
conjunction with the usual shell utilities (ls, find, etc.).
|
||||
2. Use the --xml option to specify a pre-v0.9 cross_sections.xml file.
|
||||
3. Use the --xsdir option to specify a MCNP xsdir file.
|
||||
4. Use the --xsdata option to specify a Serpent xsdata file.
|
||||
|
||||
The script does not use any extra information from cross_sections.xml/ xsdir/
|
||||
xsdata files to determine whether the nuclide is metastable. Instead, the
|
||||
--metastable argument can be used to specify whether the ZAID naming convention
|
||||
follows the NNDC data convention (1000*Z + A + 300 + 100*m), or the MCNP data
|
||||
convention (essentially the same as NNDC, except that the first metastable state
|
||||
of Am242 is 95242 and the ground state is 95642).
|
||||
|
||||
The ``openmc-ace-to-hdf5`` script has the following command-line flags:
|
||||
|
||||
-h, --help show this help message and exit
|
||||
|
||||
-d DESTINATION, --destination DESTINATION
|
||||
Directory to create new library in (default: .)
|
||||
|
||||
-m META, --metastable META
|
||||
How to interpret ZAIDs for metastable nuclides. META
|
||||
can be either 'nndc' or 'mcnp'. (default: nndc)
|
||||
|
||||
--xml XML Old-style cross_sections.xml that lists ACE libraries
|
||||
(default: None)
|
||||
|
||||
--xsdir XSDIR MCNP xsdir file that lists ACE libraries (default:
|
||||
None)
|
||||
|
||||
--xsdata XSDATA Serpent xsdata file that lists ACE libraries (default:
|
||||
None)
|
||||
|
||||
--fission_energy_release FISSION_ENERGY_RELEASE
|
||||
HDF5 file containing fission energy release data
|
||||
(default: None)
|
||||
|
||||
To use cross sections distributed with Serpent, change the <directory> element
|
||||
in the ``cross_sections_serpent.xml`` file in the root directory of the OpenMC
|
||||
distribution to the location of the Serpent cross sections. Then, either set the
|
||||
:ref:`cross_sections` in a settings.xml file or the
|
||||
:envvar:`OPENMC_CROSS_SECTIONS` environment variable to the absolute path of
|
||||
the ``cross_sections_serpent.xml``
|
||||
file.
|
||||
|
||||
Using Multi-Group Cross Sections
|
||||
--------------------------------
|
||||
|
|
@ -469,14 +516,13 @@ However, if the user has obtained or generated their own library, the user
|
|||
should set the :envvar:`OPENMC_MG_CROSS_SECTIONS` environment variable
|
||||
to the absolute path of the file library expected to used most frequently.
|
||||
|
||||
.. _NJOY: http://t2.lanl.gov/nis/codes.shtml
|
||||
.. _NJOY: http://t2.lanl.gov/nis/codes/NJOY12/
|
||||
.. _NNDC: http://www.nndc.bnl.gov/endf/b7.1/acefiles.html
|
||||
.. _NEA: http://www.oecd-nea.org
|
||||
.. _JEFF: http://www.oecd-nea.org/dbdata/jeff/
|
||||
.. _here: http://www.oecd-nea.org/dbdata/pubs/jeff312-cd.html
|
||||
.. _JEFF: https://www.oecd-nea.org/dbforms/data/eva/evatapes/jeff_32/
|
||||
.. _MCNP: http://mcnp.lanl.gov
|
||||
.. _Serpent: http://montecarlo.vtt.fi
|
||||
.. _TENDL: ftp://ftp.nrg.eu/pub/www/talys/tendl2012/tendl2012.html
|
||||
.. _TENDL: https://tendl.web.psi.ch/tendl_2015/tendl2015.html
|
||||
|
||||
--------------
|
||||
Running OpenMC
|
||||
|
|
|
|||
Loading…
Add table
Add a link
Reference in a new issue