Use openmc.data.endf.get_evaluations to read in potentially
many 318 evaluations from the heatr file. The evaluated
temperatures are pulled using the target attribute, rounded
to the nearest integer for consistency, e.g. 293.6K -> 294K.
The user can now pass a single argument indicating where output
tapes for each njoy module should be written. The default argument
is the current working directory, useful for removing any temporary
files created during testing.
The pendf, broadr, heatr, gaspr, and purr arguments can either
be booleans or strings. A boolean value indicates the module output
should be written into a file named after the module, e.g. heatr output
written to <output_dir>/heatr.
ace and xsdir can be None or strings. A value of None means the
final ace and xsdir files will be written to
<output_dir>/ace and <output_dir>/xsdir, respectively. The temporary
temperature-dependent ace and xsdir files are now always named
"ace_<temp>" and "xsdir_<temp>", as the ace and xsdir arguments can
be more than "ace" and "xsdir".
openmc.data.IncidentNeutron.from_njoy knows to use this output_dir
argument to ensure all created tape files are written into the
temporary directory or the user requested directory, if applicable.
This commit was written because simply writing the broadr, gaspr,
and purr output files to "broadr", "gaspr", and "purr" resulted
in left over files after running tests.
MT318 and MT999 for fission heating and non-fission heating
have been added to the reactions supported in
openmc/data/reaction.py and src/reaction.cpp
Changed internal language from fission-less heating to
non-fission heating
This reverts d81aaeca9 and 81c625928 in the following ways.
Fission heating data, MT318, is pulled from the heatr file
produced when running NJOY. Reaction data that is potentially
temperature dependent is set onto the IncidentNeutron object
by scaling MT318 by the ratio of the fission cross section
used in HEATR and other fission cross sections stored on the
object. This produces potentially many MT318 fission heating
KERMA coefficients on the nuclide.
The fission-less heating coefficient, MT999, if computed by
subtracting MT318 from MT301, total heating coefficients. This
reaction is marked as redundant, as it can easily be reconstructed.
MT318 is allowed to be written to the HDF5 file, while 999 is not.
When reading back in the library, MT999 is rebuilt in exactly
the same manner.
The test_heating test in tests/unit_tests/test_data_neutron.py
has been updated given the changes in this commit.
It is worth noting that the heating coefficients computed in
NJOY only contain prompt neutrons, as
k_{i,j}(E) = sigma_{i,j}(E) * (E + Q - \bar{E})
where k_{i,j} is the kerma coefficient for reaction j of material
i, sigma_{i,j} is the corresponding reaction cross section, E
is the energy of incident particle, Q is the mass-difference
Q value, and \bar{E} is the average energy of secondary particles.
Source: NJOY16 Manual on HEATR
HEATR contains the 318 fission heating data needed to
construct the energy deposition tally. Until now, the output
tape was left in the njoy temporary directory with the other
output tapes. This commit instructs openmc.data.njoy to move
the heatr output from the running directory used in njoy.run
to the working directory used in njoy.make_ace
Removed unused imports argparse and sys
In order to reduce the amount of duplicated code in building
the energy deposition tally, the logic for scoring SCORE_FISS_Q_PROMPT
and SCORE_FISS_Q_RECOV has been pulled into a new function,
score_fission_q. The code is almost entirely copied over, with some
minor cleanup operations.
Add a brief helper function get_nuc_fission_q in
src/tallies/tally_scoring.cpp that is responsible for
returning the correct fission q value for a given
score. The two allowed scores are SCORE_FISS_Q_PROMPT and
SCORE_FISS_Q_RECOV. The helper function examines the correct
dataset on the nuclide object and evaluates the function given
the incident particle energy or returns zero.
The default value of zero is returned if the fission_q_prompt_ or
fission_q_recov_ datasets are empty, or the score is not
one of the expected types.
By setting a level, the messages will be printed only if the
verbosity setting allows it. This reduces some of the noise printed
in testing. Related output from travis:
https://travis-ci.org/openmc-dev/openmc/jobs/574867362#L1666-L1683
A level of 4 was chosen, indicating this should be printed with
the OpenMC logo, headers, and results.
IncidentNeutron.from_ace now computes the fission heating
and a fission-less heating coefficient. The fission heating
is the product of the heating number and the fission cross
section, and stored as MT318.
The fission-less heating is the heating from all reactions
except fission, computed as heating_number * (total_xs - fission_xs).
The MT number for this is taken to be 999 as a temporary value.
A test is added for Am244 in test_data_neutron.py to examine the
new heating values
Remove copy method.
__contains__ and __getitem__ now take advantage of the
ordering of products. This can improve time searching
for fission products and retrieving yields.
radd and rmul methods defer to their left counterparts, e.g.
x * fy => fy * x. Return NotImplemented types if addition is
not done with another set of fission yields and multiplication
is not done with a scalar.
Improve/expand special methods for FissionYield
Remove copy method.
__contains__ and __getitem__ now take advantage of the
ordering of products. This can improve time searching
for fission products and retrieving yields.
radd and rmul methods defer to their left counterparts, e.g.
x * fy => fy * x. Return NotImplemented types if addition is
not done with another set of fission yields and multiplication
is not done with a scalar.
The FissionYieldCutoffHelper will always find a set of yields
to use for all nuclides with yield data now, rather than raise
an error. Previously, if the cutoff was outside the provided
bounds, an error was raised because there wasn't a clear set of
"fast" and "thermal" yields to use. However, this caused issues
with nuclides that are missing a set of lower yields, like
Th232 with yields at 5e5 and 6e6 per ENDF/B-VII.1 data.
Now, if the cutoff energy is outside the bounds of provided
yield data, the closet set of yields to the cutoff is taken to
be constant. These nuclides will not be tallied during the
transport routine.
Removed unit test files for test_deplete_*py that pertained to a
single integrator type only. Similarly removed tests where the
integrators are used to perform simple restart analysis.
These tests are still present, but done with two parametrized tests.
Each of the integrators has the exact same solution in the unit tests
with and without restarting,by nature of their design.
This and the near-fixture like way the tests
are structured allowed these tests to be easily parametrized provided
1) integrator, 2) reference solutions for atoms 1 and 2.
Exact results for each integrator in the unit test are contained
inside a collections.namedtuple, along with the Integrator class.
These named tuples are placed in a dictionary in
tests/dummy_operator.py that can be easily iterated over to provide
access to a name of the scheme, e.g. "predictor", and the tuple of
integrator, results for atom 1, and results for atom 2.
Exact results for atoms 1 and 2 should be provided using the depletion
matrix produced by the tests.dummy_operator.DummyOperator for two
time steps of 0.75 seconds.
Reconfigure tests/unit_tests/test_deplete_chain.py and
tests/unit_tests/test_deplete_nuclide.py to use the new
dictionary-like representation of fission yields.
Provide a module-scoped fixture that creates a temporary
directory, adds simple settings, geometry, and material files,
and initializes the openmc C API. This fixture also provides
two empty openmc.capi.Material objects to help the helpers
generate meaningful tallies.
The main tests for AveragedFissionYieldHelper and
FissionYieldCutoffHelper now go through the built tallies
and examine the filters, nuclides, and scores. A helper
function produces mocked-like tally data based on filters,
nuclides, and scores found on a tally.
Pu239 with 0.0253 eV, 500 keV, and 2 MeV yields is included in
the nuclide_bundle fixture. This provides some additional
heterogeneity in the results, as the 2 MeV data is not present
on U235.
Had to guard against some cases where it was assumed
that yield_data would be a dictionary. For the most part,
the fission yield helpers act on nuclides that have yield data
already.
The previous default state was to use an empty dictionary.
This has been removed. A value of None for nuclide.yield_data
indicates there is no yield data.
