.. _usersguide_depletion:

===========================
Depletion and Transmutation
===========================

OpenMC supports transport-coupled and transport-independent depletion, or
burnup, calculations through the :mod:`openmc.deplete` Python module. OpenMC
uses transmutation reaction rates to solve a set of transmutation equations
that determine the evolution of nuclide densities within a material. The
nuclide densities predicted at some future time are then used to determine
updated reaction rates, and the process is repeated for as many timesteps as
are requested.

The depletion module is designed such that the reaction rate solution (the
transport "operator") is completely isolated from the solution of the
transmutation equations and the method used for advancing time.

:mod:`openmc.deplete` supports multiple time-integration methods for determining
material compositions over time. Each method appears as a different class.
For example, :class:`openmc.deplete.CECMIntegrator` runs a depletion calculation
using the CE/CM algorithm (deplete over a timestep using the middle-of-step
reaction rates). An instance of :class:`~openmc.deplete.abc.TransportOperator`
is passed to one of these Integrator classes along with the timesteps and power
level::

    power = 1200.0e6  # watts
    timesteps = [10.0, 10.0, 10.0]  # days
    openmc.deplete.CECMIntegrator(op, timesteps, power, timestep_units='d').integrate()

The depletion problem is executed, and once it is done a
``depletion_results.h5`` file is written. The results can be analyzed using the
:class:`openmc.deplete.Results` class. This class has methods that allow for
easy retrieval of k-effective, nuclide concentrations, and reaction rates over
time::

    results = openmc.deplete.Results("depletion_results.h5")
    time, keff = results.get_keff()

Note that the coupling between the reaction rate solver and the transmutation
solver happens in-memory rather than by reading/writing files on disk. OpenMC
has two categories of transport operators for obtaining transmutation reaction
rates.

.. _coupled-depletion:

Transport-coupled depletion
===========================

This category of operator solves the transport equation to obtain transmutation
reaction rates. At present, the :mod:`openmc.deplete` module offers a single
transport-coupled operator, :class:`openmc.deplete.CoupledOperator` (which uses
the OpenMC transport solver), but in principle additional transport-coupled
operator classes based on other transport codes could be implemented and no
changes to the depletion solver itself would be needed. The
:class:`openmc.deplete.CoupledOperator` class requires a :class:`~openmc.Model`
instance containing material, geometry, and settings information::

    model = openmc.Model()
    ...

    op = openmc.deplete.CoupledOperator(model)

Any material that contains a fissionable nuclide is depleted by default, but
this can behavior can be changed with the :attr:`Material.depletable` attribute.

.. important::

   The volume must be specified for each material that is depleted by setting
   the :attr:`Material.volume` attribute. This is necessary in order to
   calculate the proper normalization of tally results based on the source rate.

Fixed-Source Transmutation
--------------------------

When the ``power`` or ``power_density`` argument is used for one of the
Integrator classes, it is assumed that OpenMC is running in k-eigenvalue mode,
and normalization of tally results is performed based on energy deposition. It
is also possible to run a fixed-source simulation and perform normalization
based on a known source rate. First, as with all fixed-source calculations, we
need to set the run mode::

    settings.run_mode = 'fixed source'

Additionally, all materials that you wish to deplete need to be marked as such
using the :attr:`Material.depletable` attribute::

    mat = openmc.Material()
    mat.depletable = True

When constructing the :class:`~openmc.deplete.CoupledOperator`, you should
indicate that normalization of tally results will be done based on the source
rate rather than a power or power density::

    op = openmc.deplete.CoupledOperator(model, normalization_mode='source-rate')

Finally, when creating a depletion integrator, use the ``source_rates`` argument::

    integrator = openmc.deplete.PredictorIntegrator(op, timesteps, sources_rates=...)

As with the ``power`` argument, you can provide a different source rate for each
timestep in the calculation. A zero source rate for a given timestep will result
in a decay-only step, where all reaction rates are zero.

Caveats
-------

.. _energy-deposition:

Energy Deposition
~~~~~~~~~~~~~~~~~

The default energy deposition mode, ``"fission-q"``, instructs the
:class:`~openmc.deplete.CoupledOperator` to normalize reaction rates using the
product of fission reaction rates and fission Q values taken from the depletion
chain. This approach does not consider indirect contributions to energy
deposition, such as neutron heating and energy from secondary photons. In doing
this, the energy deposited during a transport calculation will be lower than
expected. This causes the reaction rates to be over-adjusted to hit the
user-specific power, or power density, leading to an over-depletion of burnable
materials.

There are some remedies. First, the fission Q values can be directly set in a
variety of ways. This requires knowing what the total fission energy release
should be, including indirect components. Some examples are provided below::

    # use a dictionary of fission_q values
    fission_q = {"U235": 202e+6}  # energy in eV

    # create a Model object
    model = openmc.Model(geometry, settings)

    # create a modified chain and write it to a new file
    chain = openmc.deplete.Chain.from_xml("chain.xml", fission_q)
    chain.export_to_xml("chain_mod_q.xml")
    op = openmc.deplete.CoupledOperator(model, "chain_mod_q.xml")

    # alternatively, pass the modified fission Q directly to the operator
    op = openmc.deplete.CoupledOperator(model, "chain.xml",
        fission_q=fission_q)


A more complete way to model the energy deposition is to use the modified
heating reactions described in :ref:`methods_heating`. These values can be used
to normalize reaction rates instead of using the fission reaction rates with::

    op = openmc.deplete.CoupledOperator(model, "chain.xml",
        normalization_mode="energy-deposition")

These modified heating libraries can be generated by running the latest version
of :meth:`openmc.data.IncidentNeutron.from_njoy()`, and will eventually be bundled
into the distributed libraries.

Local Spectra and Repeated Materials
~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~

It is not uncommon to explicitly create a single burnable material across many
locations. From a pure transport perspective, there is nothing wrong with
creating a single 3.5 wt.% enriched fuel ``fuel_3``, and placing that fuel in
every fuel pin in an assembly or even full core problem. This certainly
expedites the model making process, but can pose issues with depletion. Under
this setup, :mod:`openmc.deplete` will deplete a single ``fuel_3`` material
using a single set of reaction rates, and produce a single new composition for
the next time step. This can be problematic if the same ``fuel_3`` is used in
very different regions of the problem.

As an example, consider a full-scale power reactor core with vacuum boundary
conditions, and with fuel pins solely composed of the same ``fuel_3`` material.
The fuel pins towards the center of the problem will surely experience a more
intense neutron flux and greater reaction rates than those towards the edge of
the domain. This indicates that the fuel in the center should be at a more
depleted state than periphery pins, at least for the fist depletion step.
However, without any other instructions, OpenMC will deplete ``fuel_3`` as a
single material, and all of the fuel pins will have an identical composition at
the next transport step.

This can be countered by instructing the operator to treat repeated instances
of the same material as a unique material definition with::

    op = openmc.deplete.CoupledOperator(model, chain_file,
        diff_burnable_mats=True)

For our example problem, this would deplete fuel on the outer region of the
problem with different reaction rates than those in the center. Materials will
be depleted corresponding to their local neutron spectra, and have unique
compositions at each transport step.  The volume of the original ``fuel_3``
material must represent the volume of **all** the ``fuel_3`` in the problem.
When creating the unique materials, this volume will be equally distributed
across all material instances.


.. note::

    This will increase the total memory usage and run time due to an increased
    number of tallies and material definitions.

Transport-independent depletion
===============================

This category of operator uses multigroup microscopic cross sections along with
multigroup flux spectra to obtain transmutation reaction rates. The cross
sections are pre-calculated, so there is no need for direct coupling between a
transport-independent operator and a transport solver. The :mod:`openmc.deplete`
module offers a single transport-independent operator,
:class:`~openmc.deplete.IndependentOperator`, and only one operator is needed
since, in theory, any transport code could calculate the multigroup microscopic
cross sections. The :class:`~openmc.deplete.IndependentOperator` class has two
constructors. The default constructor requires a :class:`openmc.Materials`
instance, a list of multigroup flux arrays, and a list of
:class:`~openmc.deplete.MicroXS` instances containing multigroup microscopic
cross sections in units of barns. This might look like the following::

    materials = openmc.Materials([m1, m2, m3])
    ...

    # Assign fluxes (generated from any code)
    flux_m1 = numpy.array([...])
    flux_m2 = numpy.array([...])
    flux_m3 = numpy.array([...])
    fluxes = [flux_m1, flux_m2, flux_m3]

    # Assign microscopic cross sections
    micro_m1 = openmc.deplete.MicroXS.from_csv('xs_m1.csv')
    micro_m2 = openmc.deplete.MicroXS.from_csv('xs_m2.csv')
    micro_m3 = openmc.deplete.MicroXS.from_csv('xs_m3.csv')
    micros = [micro_m1, micro_m2, micro_m3]

    # Create operator
    op = openmc.deplete.IndependentOperator(materials, fluxes, micros)

For more details on the :class:`~openmc.deplete.MicroXS` class, including how to
use OpenMC's transport solver to generate microscopic cross sections and fluxes
for use with :class:`~openmc.deplete.IndependentOperator`, see :ref:`micros`.

.. note::

   The same statements from :ref:`coupled-depletion` about which materials are
   depleted and the requirement for depletable materials to have a specified
   volume also apply here.

An alternate constructor,
:meth:`~openmc.deplete.IndependentOperator.from_nuclides`, accepts a volume and
dictionary of nuclide concentrations in place of the :class:`openmc.Materials`
instance. Note that while the normal constructor allows multiple materials to be
depleted with a single operator, the
:meth:`~openmc.deplete.IndependentOperator.from_nuclides` classmethod only works
for a single material::

    nuclides = {'U234': 8.92e18,
                'U235': 9.98e20,
                'U238': 2.22e22,
                'U236': 4.57e18,
                'O16': 4.64e22,
                'O17': 1.76e19}
    volume = 0.5
    op = openmc.deplete.IndependentOperator.from_nuclides(volume,
                                                          nuclides,
                                                          flux,
                                                          micro_xs,
                                                          chain_file,
                                                          nuc_units='atom/cm3')

A user can then define an integrator class as they would for a coupled
transport-depletion calculation and follow the same steps from there.

.. note::

   Ideally, multigroup cross section data should be available for every reaction
   in the depletion chain. If cross section data is not present for a nuclide in
   the depletion chain with at least one reaction, that reaction will not be
   simulated.

.. _micros:

Loading and Generating Microscopic Cross Sections
-------------------------------------------------

As mentioned above, any transport code could be used to calculate multigroup
microscopic cross sections and fluxes. The :mod:`openmc.deplete` module provides
the :class:`~openmc.deplete.MicroXS` class, which can either be instantiated
from pre-calculated cross sections in a ``.csv`` file or from data arrays
directly::

    micro_xs = MicroXS.from_csv(micro_xs_path)

    nuclides = ['U234', 'U235', 'U238']
    reactions = ['fission', '(n,gamma)']
    data = np.array([[0.1, 0.2],
                     [0.3, 0.4],
                     [0.01, 0.5]])
    micro_xs = MicroXS(data, nuclides, reactions)

.. important::

   The cross section values are assumed to be in units of barns. Make sure your
   cross sections are in the correct units before passing to a
   :class:`~openmc.deplete.IndependentOperator` object.

Additionally, a convenience function,
:func:`~openmc.deplete.get_microxs_and_flux`, can provide the needed fluxes and
cross sections using OpenMC's transport solver::

    model = openmc.Model()
    ...

    fluxes, micros = openmc.deplete.get_microxs_and_flux(model, materials)

If you are running :func:`~openmc.deplete.get_microxs_and_flux` on a cluster
where temporary files are created on a local filesystem that is not shared
across nodes, you'll need to set an environment variable pointing to a local
directoy so that each MPI process knows where to store output files used to
calculate the microscopic cross sections. In order of priority, they are
:envvar:`TMPDIR`. :envvar:`TEMP`, and :envvar:`TMP`. Users interested in further
details can read the documentation for the `tempfile
<https://docs.python.org/3/library/tempfile.html#tempfile.gettempdir>`_ module.

Caveats
-------

Reaction Rate Normalization
~~~~~~~~~~~~~~~~~~~~~~~~~~~

The :class:`~openmc.deplete.IndependentOperator` class supports two methods for
normalizing reaction rates:

.. important::

   Make sure you set the correct parameter in the :class:`openmc.abc.Integrator`
   class. Use the ``source_rates`` parameter when
   ``normalization_mode == source-rate``, and use ``power`` or ``power_density``
   when ``normalization_mode == fission-q``.

1. ``source-rate`` normalization, which assumes the ``source_rate`` provided by
   the time integrator is a flux, and obtains the reaction rates by multiplying
   the cross sections by the ``source-rate``.
2. ``fission-q`` normalization, which uses the ``power`` or ``power_density``
   provided by the time integrator to obtain normalized reaction rates by
   computing a normalization factor as the ratio of the user-specified power to
   the "observed" power based on fission reaction rates. The equation for the
   normalization factor is

   .. math::
      :label: fission-q

      f = \frac{P}{\sum\limits_m \sum\limits_i \left(Q_i N_{i,m} \sum\limits_g
      \sigma^f_{i,g,m} \phi_{g,m} \right)}

   where :math:`P` is the power, :math:`Q_i` is the fission Q value for nuclide
   :math:`i`, :math:`\sigma_{i,g,m}^f` is the microscopic fission cross section
   for nuclide :math:`i` in energy group :math:`g` for material :math:`m`,
   :math:`\phi_{g,m}` is the neutron flux in group :math:`g` for material
   :math:`m`, and :math:`N_{i,m}` is the number of atoms of nuclide :math:`i`
   for material :math:`m`. Reaction rates are then multiplied by :math:`f` so
   that the total fission power matches :math:`P`. This equation makes the same
   assumptions and issues as discussed in :ref:`energy-deposition`.
   Unfortunately, the proposed solution in that section does not apply here
   since we are decoupled from transport code. However, there is a method to
   converge to a more accurate value for flux by using substeps during time
   integration. `This paper <https://doi.org/10.1016/j.anucene.2016.05.031>`_
   provides a good discussion of this method.

.. warning::

   The accuracy of results when using ``fission-q`` is entirely dependent on
   your depletion chain. Make sure it has sufficient data to resolve the
   dynamics of your particular scenario.

Multiple Materials
~~~~~~~~~~~~~~~~~~

A transport-independent depletion simulation using ``source-rate`` normalization
will calculate reaction rates for each material independently. This can be
useful for running many different cases of a particular scenario. A
transport-independent depletion simulation using ``fission-q`` normalization
will sum the fission energy values across all materials into :math:`Q_i` in
Equation :math:numref:`fission-q`, and Equation :math:numref:`fission-q`
provides the normalization factor applied to reaction rates in each material.
This can be useful for running a scenario with multiple depletable materials
that are part of the same reactor. This behavior may change in the future.

Time integration
~~~~~~~~~~~~~~~~

The values of the microscopic cross sections passed to
:class:`openmc.deplete.IndependentOperator` are fixed for the entire depletion
simulation. This implicit assumption may produce inaccurate results for certain
scenarios.

Transfer Rates
==============

Transfer rates define removal or feed of nuclides to or from one or more
depletable materials. This can be useful to model continuous fuel reprocessing,
online fission products separation, etc.

Transfer rates are defined by calling the
:meth:`~openmc.deplete.abc.Integrator.add_transfer_rate()` method directly from
one of the Integrator classes::

    ...
    integrator = openmc.deplete.PredictorIntegrator(op, time_steps, power)
    integrator.add_transfer_rate(...)

Defining transfer rates
-----------------------

The :meth:`~openmc.deplete.abc.Integrator.add_transfer_rate()` method requires a
:class:`~openmc.Material` instance (alternatively, a material id or
the name) as the depletable material from which nuclides are processed,
a list of elements that share the same transfer rate, and a transfer rate itself.

.. caution::

   Make sure you set the transfer rate value with the right sign.
   A positive transfer rate assumes removal, while a negative one assumes feed.

The ``transfer_rate_units`` argument specifies the units for the transfer rate.
The default is `1/s`, but '1/min', '1/h', '1/d' and '1/a' are also valid
options.

For example, to define continuous removal of xenon from one material with a
removal rate value of 0.1 s\ :sup:`-1` (or a cycle time of 10 s), you'd use::

    mat1 = openmc.Material(material_id=1, name='fuel')

    ...

    integrator = openmc.deplete.PredictorIntegrator(op, time_steps, power)
    # by openmc.Material object
    integrator.add_transfer_rate(mat1, ['Xe'], 0.1)
    # or by material id
    integrator.add_transfer_rate(1, ['Xe'], 0.1)
    # or by material name
    integrator.add_transfer_rate('fuel', ['Xe'], 0.1)

Note that in this case the xenon isotopes that are removed will not be tracked.

Defining a destination material
-------------------------------

To transfer elements from one depletable material to another, the
``destination_material`` parameter needs to be passed to the
:meth:`~openmc.deplete.abc.Integrator.add_transfer_rate()` method. For example,
to transfer xenon from one material to another, you'd use::

    ...
    mat2 = openmc.Material(name='storage')

    ...

    integrator.add_transfer_rate(mat1, ['Xe'], 0.1, destination_material=mat2)

Comparing to Other Codes
========================

Comparing depletion results from OpenMC with those from another code, such as
MCNP or Serpent, requires more than constructing equivalent transport models.
At each depletion step, differences in the transport solution, nuclear data,
reaction rate normalization, and numerical integration can all affect the
result. Small differences can also accumulate over successive depletion steps.

For a meaningful comparison, align as many of the following inputs and methods
as possible:

- Geometry and material definitions and associated physical properties such as
  temperature
- Neutron cross section library (e.g., ENDF/B-VIII.0)
- Treatment of thermal scattering and unresolved resonance probability tables
- Neutron reactions accounted for in the depletion chain
- Decay data in the depletion chain
- Isomeric branching ratios for reactions in the depletion chain
- Fission product yields in the depletion chain
- Fission product yield interpolation method
  (``CoupledOperator(fission_yield_mode=...)``)
- Reaction rate normalization, including fission Q values
  (``CoupledOperator(fission_q=...)``)
- Depletion integration method (``PredictorIntegrator``, ``CECMIntegrator``,
  etc.) and time-step sizes

When comparing to codes that use ACE format cross sections, it is recommended to
directly convert the ACE files to HDF5 format using functionality from the
:mod:`openmc.data` module (see :ref:`create_xs_library`). Some of the
LANL-distributed ACE libraries used with MCNP have also been converted to HDF5
format and are available for download at https://openmc.org/data.

Even after these choices have been aligned, exact agreement should not be
expected. Codes may use different approximations or numerical methods that
cannot be configured identically. When investigating a discrepancy, first
compare transport results and one-group reaction rates at the initial time, then
compare changes over subsequent timesteps.
