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OpenMC/tests/unit_tests/test_deplete_fission_yields.py

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Add ConstantFissionYieldHelper concrete class Given a requested energy, will take fission yields from that energy on all nuclides and hold them as constant throughout the simulation. If the requested energy is not found for a specific nuclide, then the closest energy is used. The default is to take the thermal 0.0253 eV yields. This is now the helper attached to the Operator. The user can select what energy of yields they would like to use.
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"""Test the FissionYieldHelpers"""
Improve testing for FY helpers using C API 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.
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import os
Add FissionYieldCutoffHelper for weighting yields by a cutoff energy This helper performs the following tasks to weight fission yields: 1) Determine a set of "fast" and "thermal" yields based on a user specified cutoff energy, 2) Tally fission rates above and below the cutoff, and 3) Compute the effective yield by weighting the fast and thermal yields by the fraction of fast and thermal fissions for all nuclides in all burnable materials. The user is allowed to specify the cutoff energy and energies preferred for the fast and thermal yields. By default, the cutoff is 112 eV, chosen as it is the logarithmic mean of 0.0253 eV and 500 keV. These two values are the default energies for thermal and fast yields. Tests are added to check for failure in constructing this helper (thermal energy > cutoff, non-real values passed, etc.), selection of yields given a range of user data, and a proxy class that emulates a variety of thermal and fast splits to check the eventual computed yield libraries.
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from collections import namedtuple
from unittest.mock import Mock
Improve testing for FY helpers using C API 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.
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import bisect
Add FissionYieldCutoffHelper for weighting yields by a cutoff energy This helper performs the following tasks to weight fission yields: 1) Determine a set of "fast" and "thermal" yields based on a user specified cutoff energy, 2) Tally fission rates above and below the cutoff, and 3) Compute the effective yield by weighting the fast and thermal yields by the fraction of fast and thermal fissions for all nuclides in all burnable materials. The user is allowed to specify the cutoff energy and energies preferred for the fast and thermal yields. By default, the cutoff is 112 eV, chosen as it is the logarithmic mean of 0.0253 eV and 500 keV. These two values are the default energies for thermal and fast yields. Tests are added to check for failure in constructing this helper (thermal energy > cutoff, non-real values passed, etc.), selection of yields given a range of user data, and a proxy class that emulates a variety of thermal and fast splits to check the eventual computed yield libraries.
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import pytest
Use python API for setup in test_deplete_fission_yields
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import numpy as np
import openmc
Rename openmc.capi -> openmc.lib
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from openmc import lib
Add FissionYieldCutoffHelper for weighting yields by a cutoff energy This helper performs the following tasks to weight fission yields: 1) Determine a set of "fast" and "thermal" yields based on a user specified cutoff energy, 2) Tally fission rates above and below the cutoff, and 3) Compute the effective yield by weighting the fast and thermal yields by the fraction of fast and thermal fissions for all nuclides in all burnable materials. The user is allowed to specify the cutoff energy and energies preferred for the fast and thermal yields. By default, the cutoff is 112 eV, chosen as it is the logarithmic mean of 0.0253 eV and 500 keV. These two values are the default energies for thermal and fast yields. Tests are added to check for failure in constructing this helper (thermal energy > cutoff, non-real values passed, etc.), selection of yields given a range of user data, and a proxy class that emulates a variety of thermal and fast splits to check the eventual computed yield libraries.
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from openmc.deplete.nuclide import Nuclide, FissionYieldDistribution
from openmc.deplete.helpers import (
Add AveragedFissionYieldHelper operator helper Computes the effective fission yields for nuclides with multiple sets of yields by 1) Tallying fission rate and energy-weighted fission rate 2) Determining average energy at which fission events occur 3) Interpolating [lin-lin] between adjacent fission yields based on this average energy Uses a second tally with an EnergyFunctionFilter to tally E * sigma_f * phi. Added unit tests with the proxy-style class that doesn't use the C API to generate tallies but tests all other aspects of the implementation. Tests cover three possible average energies (below minimum supplied yield energy, above max supplied yield energy, and between two) The Operator uses fission_yield_mode="average" to select this helper. The helper also has a from_operator class method.
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FissionYieldCutoffHelper, ConstantFissionYieldHelper,
AveragedFissionYieldHelper)
Improve testing for FY helpers using C API 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.
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@pytest.fixture(scope="module")
def materials(tmpdir_factory):
"""Use C API to construct realistic materials for testing tallies"""
Rename openmc.capi -> openmc.lib
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tmpdir = tmpdir_factory.mktemp("lib")
Improve testing for FY helpers using C API 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.
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orig = tmpdir.chdir()
Use python API for setup in test_deplete_fission_yields
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# Create proxy problem to please openmc
mfuel = openmc.Material(name="test_fuel")
mfuel.volume = 1.0
for nuclide in ["U235", "U238", "Xe135", "Pu239"]:
mfuel.add_nuclide(nuclide, 1.0)
openmc.Materials([mfuel]).export_to_xml()
# Geometry
Add RectangularPrism and HexagonalPrism composite surfaces (#2739) Co-authored-by: Patrick Shriwise <pshriwise@gmail.com>
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box = openmc.model.RectangularPrism(1.0, 1.0, boundary_type="reflective")
cell = openmc.Cell(fill=mfuel, region=-box)
Use python API for setup in test_deplete_fission_yields
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root = openmc.Universe(cells=[cell])
openmc.Geometry(root).export_to_xml()
# settings
settings = openmc.Settings()
settings.particles = 100
settings.inactive = 0
settings.batches = 10
settings.verbosity = 1
settings.export_to_xml()
Improve testing for FY helpers using C API 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.
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try:
Rename openmc.capi -> openmc.lib
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with lib.run_in_memory():
yield [lib.Material(), lib.Material()]
Improve testing for FY helpers using C API 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.
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finally:
Pass strings to os functions in test_deplete_fission_yields os.remove and os.chdir support path-like objects only for python 3.6+
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for file_path in ("settings.xml", "geometry.xml", "materials.xml",
"summary.h5"):
Resolve a few TODOs related to Python 3.6
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os.remove(tmpdir / file_path)
Improve testing for FY helpers using C API 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.
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orig.chdir()
Resolve a few TODOs related to Python 3.6
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os.rmdir(tmpdir)
Improve testing for FY helpers using C API 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.
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def proxy_tally_data(tally, fill=None):
"""Construct an empty matrix built from a C tally
Use tally mean in helpers rather than direct access to results
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The shape of tally.mean will be ``(n_bins, n_nuc * n_scores)``
Improve testing for FY helpers using C API 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.
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"""
n_nucs = max(len(tally.nuclides), 1)
n_scores = max(len(tally.scores), 1)
n_bins = 1
for tfilter in tally.filters:
if not hasattr(tfilter, "bins"):
continue
this_bins = len(tfilter.bins)
Rename openmc.capi -> openmc.lib
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if isinstance(tfilter, lib.EnergyFilter):
Improve testing for FY helpers using C API 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.
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this_bins -= 1
n_bins *= max(this_bins, 1)
Use tally mean in helpers rather than direct access to results
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data = np.empty((n_bins, n_nucs * n_scores))
Improve testing for FY helpers using C API 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.
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if fill is not None:
data.fill(fill)
return data
Add FissionYieldCutoffHelper for weighting yields by a cutoff energy This helper performs the following tasks to weight fission yields: 1) Determine a set of "fast" and "thermal" yields based on a user specified cutoff energy, 2) Tally fission rates above and below the cutoff, and 3) Compute the effective yield by weighting the fast and thermal yields by the fraction of fast and thermal fissions for all nuclides in all burnable materials. The user is allowed to specify the cutoff energy and energies preferred for the fast and thermal yields. By default, the cutoff is 112 eV, chosen as it is the logarithmic mean of 0.0253 eV and 500 keV. These two values are the default energies for thermal and fast yields. Tests are added to check for failure in constructing this helper (thermal energy > cutoff, non-real values passed, etc.), selection of yields given a range of user data, and a proxy class that emulates a variety of thermal and fast splits to check the eventual computed yield libraries.
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@pytest.fixture(scope="module")
def nuclide_bundle():
u5yield_dict = {
0.0253: {"Xe135": 7.85e-4, "Gd155": 4.08e-12, "Sm149": 1.71e-12},
5.0e5: {"Xe135": 7.85e-4, "Sm149": 1.71e-12},
1.40e7: {"Xe135": 4.54e-3, "Gd155": 5.83e-8}}
Support passing nuclide name to Nuclide objects Allow the creation of a Nuclide with >>> n = Nuclide("U235")
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u235 = Nuclide("U235")
Use dictionary to create FissionYieldDistribution Removes the FissionYieldDistribution.from_dict method as the most natural way to create such a distribution is with a dictionary.
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u235.yield_data = FissionYieldDistribution(u5yield_dict)
Add FissionYieldCutoffHelper for weighting yields by a cutoff energy This helper performs the following tasks to weight fission yields: 1) Determine a set of "fast" and "thermal" yields based on a user specified cutoff energy, 2) Tally fission rates above and below the cutoff, and 3) Compute the effective yield by weighting the fast and thermal yields by the fraction of fast and thermal fissions for all nuclides in all burnable materials. The user is allowed to specify the cutoff energy and energies preferred for the fast and thermal yields. By default, the cutoff is 112 eV, chosen as it is the logarithmic mean of 0.0253 eV and 500 keV. These two values are the default energies for thermal and fast yields. Tests are added to check for failure in constructing this helper (thermal energy > cutoff, non-real values passed, etc.), selection of yields given a range of user data, and a proxy class that emulates a variety of thermal and fast splits to check the eventual computed yield libraries.
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u8yield_dict = {5.00e5: {"Xe135": 1.12e-3, "Gd155": 1.32e-12}}
Support passing nuclide name to Nuclide objects Allow the creation of a Nuclide with >>> n = Nuclide("U235")
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u238 = Nuclide("U238")
Use dictionary to create FissionYieldDistribution Removes the FissionYieldDistribution.from_dict method as the most natural way to create such a distribution is with a dictionary.
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u238.yield_data = FissionYieldDistribution(u8yield_dict)
Add FissionYieldCutoffHelper for weighting yields by a cutoff energy This helper performs the following tasks to weight fission yields: 1) Determine a set of "fast" and "thermal" yields based on a user specified cutoff energy, 2) Tally fission rates above and below the cutoff, and 3) Compute the effective yield by weighting the fast and thermal yields by the fraction of fast and thermal fissions for all nuclides in all burnable materials. The user is allowed to specify the cutoff energy and energies preferred for the fast and thermal yields. By default, the cutoff is 112 eV, chosen as it is the logarithmic mean of 0.0253 eV and 500 keV. These two values are the default energies for thermal and fast yields. Tests are added to check for failure in constructing this helper (thermal energy > cutoff, non-real values passed, etc.), selection of yields given a range of user data, and a proxy class that emulates a variety of thermal and fast splits to check the eventual computed yield libraries.
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Support passing nuclide name to Nuclide objects Allow the creation of a Nuclide with >>> n = Nuclide("U235")
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xe135 = Nuclide("Xe135")
Add FissionYieldCutoffHelper for weighting yields by a cutoff energy This helper performs the following tasks to weight fission yields: 1) Determine a set of "fast" and "thermal" yields based on a user specified cutoff energy, 2) Tally fission rates above and below the cutoff, and 3) Compute the effective yield by weighting the fast and thermal yields by the fraction of fast and thermal fissions for all nuclides in all burnable materials. The user is allowed to specify the cutoff energy and energies preferred for the fast and thermal yields. By default, the cutoff is 112 eV, chosen as it is the logarithmic mean of 0.0253 eV and 500 keV. These two values are the default energies for thermal and fast yields. Tests are added to check for failure in constructing this helper (thermal energy > cutoff, non-real values passed, etc.), selection of yields given a range of user data, and a proxy class that emulates a variety of thermal and fast splits to check the eventual computed yield libraries.
2019-08-13 16:38:24 -05:00
Improve testing for FY helpers using C API 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.
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pu239 = Nuclide("Pu239")
pu239.yield_data = FissionYieldDistribution({
5.0e5: {"Xe135": 6.14e-3, "Sm149": 9.429e-10, "Gd155": 5.24e-9},
2e6: {"Xe135": 6.15e-3, "Sm149": 9.42e-10, "Gd155": 5.29e-9}})
NuclideBundle = namedtuple("NuclideBundle", "u235 u238 xe135 pu239")
return NuclideBundle(u235, u238, xe135, pu239)
Fix wonky layout in test_deplete_fission_yields
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@pytest.mark.parametrize(
Improve testing for FY helpers using C API 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.
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"input_energy, yield_energy",
Fix wonky layout in test_deplete_fission_yields
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((0.0253, 0.0253), (0.01, 0.0253), (4e5, 5e5)))
Improve testing for FY helpers using C API 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.
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def test_constant_helper(nuclide_bundle, input_energy, yield_energy):
Fix wonky layout in test_deplete_fission_yields
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helper = ConstantFissionYieldHelper(nuclide_bundle, energy=input_energy)
Add FissionYieldCutoffHelper for weighting yields by a cutoff energy This helper performs the following tasks to weight fission yields: 1) Determine a set of "fast" and "thermal" yields based on a user specified cutoff energy, 2) Tally fission rates above and below the cutoff, and 3) Compute the effective yield by weighting the fast and thermal yields by the fraction of fast and thermal fissions for all nuclides in all burnable materials. The user is allowed to specify the cutoff energy and energies preferred for the fast and thermal yields. By default, the cutoff is 112 eV, chosen as it is the logarithmic mean of 0.0253 eV and 500 keV. These two values are the default energies for thermal and fast yields. Tests are added to check for failure in constructing this helper (thermal energy > cutoff, non-real values passed, etc.), selection of yields given a range of user data, and a proxy class that emulates a variety of thermal and fast splits to check the eventual computed yield libraries.
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assert helper.energy == input_energy
assert helper.constant_yields == {
Improve testing for FY helpers using C API 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.
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"U235": nuclide_bundle.u235.yield_data[yield_energy],
Use constant FY if cutoff energy outside provided yields 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.
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"U238": nuclide_bundle.u238.yield_data[5.00e5],
"Pu239": nuclide_bundle.pu239.yield_data[5e5]}
Add FissionYieldCutoffHelper for weighting yields by a cutoff energy This helper performs the following tasks to weight fission yields: 1) Determine a set of "fast" and "thermal" yields based on a user specified cutoff energy, 2) Tally fission rates above and below the cutoff, and 3) Compute the effective yield by weighting the fast and thermal yields by the fraction of fast and thermal fissions for all nuclides in all burnable materials. The user is allowed to specify the cutoff energy and energies preferred for the fast and thermal yields. By default, the cutoff is 112 eV, chosen as it is the logarithmic mean of 0.0253 eV and 500 keV. These two values are the default energies for thermal and fast yields. Tests are added to check for failure in constructing this helper (thermal energy > cutoff, non-real values passed, etc.), selection of yields given a range of user data, and a proxy class that emulates a variety of thermal and fast splits to check the eventual computed yield libraries.
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assert helper.constant_yields == helper.weighted_yields(1)
def test_cutoff_construction(nuclide_bundle):
Improve testing for FY helpers using C API 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.
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u235 = nuclide_bundle.u235
u238 = nuclide_bundle.u238
pu239 = nuclide_bundle.pu239
Add FissionYieldCutoffHelper for weighting yields by a cutoff energy This helper performs the following tasks to weight fission yields: 1) Determine a set of "fast" and "thermal" yields based on a user specified cutoff energy, 2) Tally fission rates above and below the cutoff, and 3) Compute the effective yield by weighting the fast and thermal yields by the fraction of fast and thermal fissions for all nuclides in all burnable materials. The user is allowed to specify the cutoff energy and energies preferred for the fast and thermal yields. By default, the cutoff is 112 eV, chosen as it is the logarithmic mean of 0.0253 eV and 500 keV. These two values are the default energies for thermal and fast yields. Tests are added to check for failure in constructing this helper (thermal energy > cutoff, non-real values passed, etc.), selection of yields given a range of user data, and a proxy class that emulates a variety of thermal and fast splits to check the eventual computed yield libraries.
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# defaults
Fix wonky layout in test_deplete_fission_yields
2019-08-14 15:15:47 -05:00
helper = FissionYieldCutoffHelper(nuclide_bundle, 1)
Add FissionYieldCutoffHelper for weighting yields by a cutoff energy This helper performs the following tasks to weight fission yields: 1) Determine a set of "fast" and "thermal" yields based on a user specified cutoff energy, 2) Tally fission rates above and below the cutoff, and 3) Compute the effective yield by weighting the fast and thermal yields by the fraction of fast and thermal fissions for all nuclides in all burnable materials. The user is allowed to specify the cutoff energy and energies preferred for the fast and thermal yields. By default, the cutoff is 112 eV, chosen as it is the logarithmic mean of 0.0253 eV and 500 keV. These two values are the default energies for thermal and fast yields. Tests are added to check for failure in constructing this helper (thermal energy > cutoff, non-real values passed, etc.), selection of yields given a range of user data, and a proxy class that emulates a variety of thermal and fast splits to check the eventual computed yield libraries.
2019-08-13 16:38:24 -05:00
assert helper.constant_yields == {
Use constant FY if cutoff energy outside provided yields 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.
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"U238": u238.yield_data[5.0e5],
Improve testing for FY helpers using C API 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.
2019-08-21 17:20:19 -05:00
"Pu239": pu239.yield_data[5e5]}
Use constant FY if cutoff energy outside provided yields 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.
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assert helper.thermal_yields == {"U235": u235.yield_data[0.0253]}
assert helper.fast_yields == {"U235": u235.yield_data[5e5]}
Improve testing for FY helpers using C API 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.
2019-08-21 17:20:19 -05:00
Add FissionYieldCutoffHelper for weighting yields by a cutoff energy This helper performs the following tasks to weight fission yields: 1) Determine a set of "fast" and "thermal" yields based on a user specified cutoff energy, 2) Tally fission rates above and below the cutoff, and 3) Compute the effective yield by weighting the fast and thermal yields by the fraction of fast and thermal fissions for all nuclides in all burnable materials. The user is allowed to specify the cutoff energy and energies preferred for the fast and thermal yields. By default, the cutoff is 112 eV, chosen as it is the logarithmic mean of 0.0253 eV and 500 keV. These two values are the default energies for thermal and fast yields. Tests are added to check for failure in constructing this helper (thermal energy > cutoff, non-real values passed, etc.), selection of yields given a range of user data, and a proxy class that emulates a variety of thermal and fast splits to check the eventual computed yield libraries.
2019-08-13 16:38:24 -05:00
# use 14 MeV yields
Fix wonky layout in test_deplete_fission_yields
2019-08-14 15:15:47 -05:00
helper = FissionYieldCutoffHelper(nuclide_bundle, 1, fast_energy=14e6)
Add FissionYieldCutoffHelper for weighting yields by a cutoff energy This helper performs the following tasks to weight fission yields: 1) Determine a set of "fast" and "thermal" yields based on a user specified cutoff energy, 2) Tally fission rates above and below the cutoff, and 3) Compute the effective yield by weighting the fast and thermal yields by the fraction of fast and thermal fissions for all nuclides in all burnable materials. The user is allowed to specify the cutoff energy and energies preferred for the fast and thermal yields. By default, the cutoff is 112 eV, chosen as it is the logarithmic mean of 0.0253 eV and 500 keV. These two values are the default energies for thermal and fast yields. Tests are added to check for failure in constructing this helper (thermal energy > cutoff, non-real values passed, etc.), selection of yields given a range of user data, and a proxy class that emulates a variety of thermal and fast splits to check the eventual computed yield libraries.
2019-08-13 16:38:24 -05:00
assert helper.constant_yields == {
Use constant FY if cutoff energy outside provided yields 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.
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"U238": u238.yield_data[5.0e5],
"Pu239": pu239.yield_data[5e5]}
assert helper.thermal_yields == {"U235": u235.yield_data[0.0253]}
assert helper.fast_yields == {"U235": u235.yield_data[14e6]}
Improve testing for FY helpers using C API 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.
2019-08-21 17:20:19 -05:00
Add FissionYieldCutoffHelper for weighting yields by a cutoff energy This helper performs the following tasks to weight fission yields: 1) Determine a set of "fast" and "thermal" yields based on a user specified cutoff energy, 2) Tally fission rates above and below the cutoff, and 3) Compute the effective yield by weighting the fast and thermal yields by the fraction of fast and thermal fissions for all nuclides in all burnable materials. The user is allowed to specify the cutoff energy and energies preferred for the fast and thermal yields. By default, the cutoff is 112 eV, chosen as it is the logarithmic mean of 0.0253 eV and 500 keV. These two values are the default energies for thermal and fast yields. Tests are added to check for failure in constructing this helper (thermal energy > cutoff, non-real values passed, etc.), selection of yields given a range of user data, and a proxy class that emulates a variety of thermal and fast splits to check the eventual computed yield libraries.
2019-08-13 16:38:24 -05:00
# specify missing thermal yields -> use 0.0253
Fix wonky layout in test_deplete_fission_yields
2019-08-14 15:15:47 -05:00
helper = FissionYieldCutoffHelper(nuclide_bundle, 1, thermal_energy=1)
Use constant FY if cutoff energy outside provided yields 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.
2019-08-22 11:24:39 -05:00
assert helper.thermal_yields == {"U235": u235.yield_data[0.0253]}
assert helper.fast_yields == {"U235": u235.yield_data[5e5]}
Improve testing for FY helpers using C API 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.
2019-08-21 17:20:19 -05:00
Add FissionYieldCutoffHelper for weighting yields by a cutoff energy This helper performs the following tasks to weight fission yields: 1) Determine a set of "fast" and "thermal" yields based on a user specified cutoff energy, 2) Tally fission rates above and below the cutoff, and 3) Compute the effective yield by weighting the fast and thermal yields by the fraction of fast and thermal fissions for all nuclides in all burnable materials. The user is allowed to specify the cutoff energy and energies preferred for the fast and thermal yields. By default, the cutoff is 112 eV, chosen as it is the logarithmic mean of 0.0253 eV and 500 keV. These two values are the default energies for thermal and fast yields. Tests are added to check for failure in constructing this helper (thermal energy > cutoff, non-real values passed, etc.), selection of yields given a range of user data, and a proxy class that emulates a variety of thermal and fast splits to check the eventual computed yield libraries.
2019-08-13 16:38:24 -05:00
# request missing fast yields -> use epithermal
Fix wonky layout in test_deplete_fission_yields
2019-08-14 15:15:47 -05:00
helper = FissionYieldCutoffHelper(nuclide_bundle, 1, fast_energy=1e4)
Use constant FY if cutoff energy outside provided yields 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.
2019-08-22 11:24:39 -05:00
assert helper.thermal_yields == {"U235": u235.yield_data[0.0253]}
assert helper.fast_yields == {"U235": u235.yield_data[5e5]}
# higher cutoff energy -> obtain fast and "faster" yields
helper = FissionYieldCutoffHelper(nuclide_bundle, 1, cutoff=1e6,
thermal_energy=5e5, fast_energy=14e6)
assert helper.constant_yields == {"U238": u238.yield_data[5e5]}
Add FissionYieldCutoffHelper for weighting yields by a cutoff energy This helper performs the following tasks to weight fission yields: 1) Determine a set of "fast" and "thermal" yields based on a user specified cutoff energy, 2) Tally fission rates above and below the cutoff, and 3) Compute the effective yield by weighting the fast and thermal yields by the fraction of fast and thermal fissions for all nuclides in all burnable materials. The user is allowed to specify the cutoff energy and energies preferred for the fast and thermal yields. By default, the cutoff is 112 eV, chosen as it is the logarithmic mean of 0.0253 eV and 500 keV. These two values are the default energies for thermal and fast yields. Tests are added to check for failure in constructing this helper (thermal energy > cutoff, non-real values passed, etc.), selection of yields given a range of user data, and a proxy class that emulates a variety of thermal and fast splits to check the eventual computed yield libraries.
2019-08-13 16:38:24 -05:00
assert helper.thermal_yields == {
Use constant FY if cutoff energy outside provided yields 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.
2019-08-22 11:24:39 -05:00
"U235": u235.yield_data[5e5], "Pu239": pu239.yield_data[5e5]}
Improve testing for FY helpers using C API 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.
2019-08-21 17:20:19 -05:00
assert helper.fast_yields == {
Use constant FY if cutoff energy outside provided yields 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.
2019-08-22 11:24:39 -05:00
"U235": u235.yield_data[14e6], "Pu239": pu239.yield_data[2e6]}
# test super low and super high cutoff energies
helper = FissionYieldCutoffHelper(
nuclide_bundle, 1, thermal_energy=0.001, cutoff=0.002)
assert helper.fast_yields == {}
assert helper.thermal_yields == {}
assert helper.constant_yields == {
"U235": u235.yield_data[0.0253], "U238": u238.yield_data[5e5],
Improve testing for FY helpers using C API 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.
2019-08-21 17:20:19 -05:00
"Pu239": pu239.yield_data[5e5]}
Use constant FY if cutoff energy outside provided yields 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.
2019-08-22 11:24:39 -05:00
helper = FissionYieldCutoffHelper(
nuclide_bundle, 1, cutoff=15e6, fast_energy=17e6)
assert helper.thermal_yields == {}
assert helper.fast_yields == {}
assert helper.constant_yields == {
"U235": u235.yield_data[14e6], "U238": u238.yield_data[5e5],
"Pu239": pu239.yield_data[2e6]}
Add FissionYieldCutoffHelper for weighting yields by a cutoff energy This helper performs the following tasks to weight fission yields: 1) Determine a set of "fast" and "thermal" yields based on a user specified cutoff energy, 2) Tally fission rates above and below the cutoff, and 3) Compute the effective yield by weighting the fast and thermal yields by the fraction of fast and thermal fissions for all nuclides in all burnable materials. The user is allowed to specify the cutoff energy and energies preferred for the fast and thermal yields. By default, the cutoff is 112 eV, chosen as it is the logarithmic mean of 0.0253 eV and 500 keV. These two values are the default energies for thermal and fast yields. Tests are added to check for failure in constructing this helper (thermal energy > cutoff, non-real values passed, etc.), selection of yields given a range of user data, and a proxy class that emulates a variety of thermal and fast splits to check the eventual computed yield libraries.
2019-08-13 16:38:24 -05:00
@pytest.mark.parametrize("key", ("cutoff", "thermal_energy", "fast_energy"))
def test_cutoff_failure(key):
with pytest.raises(TypeError, match=key):
FissionYieldCutoffHelper(None, None, **{key: None})
with pytest.raises(ValueError, match=key):
FissionYieldCutoffHelper(None, None, **{key: -1})
# emulate some split between fast and thermal U235 fissions
@pytest.mark.parametrize("therm_frac", (0.5, 0.2, 0.8))
Improve testing for FY helpers using C API 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.
2019-08-21 17:20:19 -05:00
def test_cutoff_helper(materials, nuclide_bundle, therm_frac):
Use constant FY if cutoff energy outside provided yields 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.
2019-08-22 11:24:39 -05:00
helper = FissionYieldCutoffHelper(nuclide_bundle, len(materials),
cutoff=1e6, fast_energy=14e6)
Improve testing for FY helpers using C API 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.
2019-08-21 17:20:19 -05:00
helper.generate_tallies(materials, [0])
Add FissionYieldCutoffHelper for weighting yields by a cutoff energy This helper performs the following tasks to weight fission yields: 1) Determine a set of "fast" and "thermal" yields based on a user specified cutoff energy, 2) Tally fission rates above and below the cutoff, and 3) Compute the effective yield by weighting the fast and thermal yields by the fraction of fast and thermal fissions for all nuclides in all burnable materials. The user is allowed to specify the cutoff energy and energies preferred for the fast and thermal yields. By default, the cutoff is 112 eV, chosen as it is the logarithmic mean of 0.0253 eV and 500 keV. These two values are the default energies for thermal and fast yields. Tests are added to check for failure in constructing this helper (thermal energy > cutoff, non-real values passed, etc.), selection of yields given a range of user data, and a proxy class that emulates a variety of thermal and fast splits to check the eventual computed yield libraries.
2019-08-13 16:38:24 -05:00
non_zero_nucs = [n.name for n in nuclide_bundle]
Improve testing for FY helpers using C API 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.
2019-08-21 17:20:19 -05:00
tally_nucs = helper.update_tally_nuclides(non_zero_nucs)
Clean up internals for AveragedFissionYieldHelper Respond to some other reviewer comments for #1313
2019-08-30 09:36:28 -05:00
assert tally_nucs == ["Pu239", "U235"]
Improve testing for FY helpers using C API 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.
2019-08-21 17:20:19 -05:00
# Check tallies
fission_tally = helper._fission_rate_tally
assert fission_tally is not None
filters = fission_tally.filters
assert len(filters) == 2
Rename openmc.capi -> openmc.lib
2019-09-05 07:31:13 -05:00
assert isinstance(filters[0], lib.MaterialFilter)
Improve testing for FY helpers using C API 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.
2019-08-21 17:20:19 -05:00
assert len(filters[0].bins) == len(materials)
Rename openmc.capi -> openmc.lib
2019-09-05 07:31:13 -05:00
assert isinstance(filters[1], lib.EnergyFilter)
Improve testing for FY helpers using C API 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.
2019-08-21 17:20:19 -05:00
# lower, cutoff, and upper energy
assert len(filters[1].bins) == 3
Add FissionYieldCutoffHelper for weighting yields by a cutoff energy This helper performs the following tasks to weight fission yields: 1) Determine a set of "fast" and "thermal" yields based on a user specified cutoff energy, 2) Tally fission rates above and below the cutoff, and 3) Compute the effective yield by weighting the fast and thermal yields by the fraction of fast and thermal fissions for all nuclides in all burnable materials. The user is allowed to specify the cutoff energy and energies preferred for the fast and thermal yields. By default, the cutoff is 112 eV, chosen as it is the logarithmic mean of 0.0253 eV and 500 keV. These two values are the default energies for thermal and fast yields. Tests are added to check for failure in constructing this helper (thermal energy > cutoff, non-real values passed, etc.), selection of yields given a range of user data, and a proxy class that emulates a variety of thermal and fast splits to check the eventual computed yield libraries.
2019-08-13 16:38:24 -05:00
# Emulate building tallies
# material x energy, tallied_nuclides, 3
Improve testing for FY helpers using C API 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.
2019-08-21 17:20:19 -05:00
tally_data = proxy_tally_data(fission_tally)
helper._fission_rate_tally = Mock()
helper_flux = 1e6
Use tally mean in helpers rather than direct access to results
2020-07-22 22:50:05 -05:00
tally_data[0] = therm_frac * helper_flux
tally_data[1] = (1 - therm_frac) * helper_flux
helper._fission_rate_tally.mean = tally_data
Add FissionYieldCutoffHelper for weighting yields by a cutoff energy This helper performs the following tasks to weight fission yields: 1) Determine a set of "fast" and "thermal" yields based on a user specified cutoff energy, 2) Tally fission rates above and below the cutoff, and 3) Compute the effective yield by weighting the fast and thermal yields by the fraction of fast and thermal fissions for all nuclides in all burnable materials. The user is allowed to specify the cutoff energy and energies preferred for the fast and thermal yields. By default, the cutoff is 112 eV, chosen as it is the logarithmic mean of 0.0253 eV and 500 keV. These two values are the default energies for thermal and fast yields. Tests are added to check for failure in constructing this helper (thermal energy > cutoff, non-real values passed, etc.), selection of yields given a range of user data, and a proxy class that emulates a variety of thermal and fast splits to check the eventual computed yield libraries.
2019-08-13 16:38:24 -05:00
Improve testing for FY helpers using C API 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.
2019-08-21 17:20:19 -05:00
helper.unpack()
Add FissionYieldCutoffHelper for weighting yields by a cutoff energy This helper performs the following tasks to weight fission yields: 1) Determine a set of "fast" and "thermal" yields based on a user specified cutoff energy, 2) Tally fission rates above and below the cutoff, and 3) Compute the effective yield by weighting the fast and thermal yields by the fraction of fast and thermal fissions for all nuclides in all burnable materials. The user is allowed to specify the cutoff energy and energies preferred for the fast and thermal yields. By default, the cutoff is 112 eV, chosen as it is the logarithmic mean of 0.0253 eV and 500 keV. These two values are the default energies for thermal and fast yields. Tests are added to check for failure in constructing this helper (thermal energy > cutoff, non-real values passed, etc.), selection of yields given a range of user data, and a proxy class that emulates a variety of thermal and fast splits to check the eventual computed yield libraries.
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# expected results of shape (n_mats, 2, n_tnucs)
Use python API for setup in test_deplete_fission_yields
2019-08-30 09:55:58 -05:00
expected_results = np.empty((1, 2, len(tally_nucs)))
Add FissionYieldCutoffHelper for weighting yields by a cutoff energy This helper performs the following tasks to weight fission yields: 1) Determine a set of "fast" and "thermal" yields based on a user specified cutoff energy, 2) Tally fission rates above and below the cutoff, and 3) Compute the effective yield by weighting the fast and thermal yields by the fraction of fast and thermal fissions for all nuclides in all burnable materials. The user is allowed to specify the cutoff energy and energies preferred for the fast and thermal yields. By default, the cutoff is 112 eV, chosen as it is the logarithmic mean of 0.0253 eV and 500 keV. These two values are the default energies for thermal and fast yields. Tests are added to check for failure in constructing this helper (thermal energy > cutoff, non-real values passed, etc.), selection of yields given a range of user data, and a proxy class that emulates a variety of thermal and fast splits to check the eventual computed yield libraries.
2019-08-13 16:38:24 -05:00
expected_results[:, 0] = therm_frac
Fix wonky layout in test_deplete_fission_yields
2019-08-14 15:15:47 -05:00
expected_results[:, 1] = 1 - therm_frac
Improve testing for FY helpers using C API 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.
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assert helper.results == pytest.approx(expected_results)
Add FissionYieldCutoffHelper for weighting yields by a cutoff energy This helper performs the following tasks to weight fission yields: 1) Determine a set of "fast" and "thermal" yields based on a user specified cutoff energy, 2) Tally fission rates above and below the cutoff, and 3) Compute the effective yield by weighting the fast and thermal yields by the fraction of fast and thermal fissions for all nuclides in all burnable materials. The user is allowed to specify the cutoff energy and energies preferred for the fast and thermal yields. By default, the cutoff is 112 eV, chosen as it is the logarithmic mean of 0.0253 eV and 500 keV. These two values are the default energies for thermal and fast yields. Tests are added to check for failure in constructing this helper (thermal energy > cutoff, non-real values passed, etc.), selection of yields given a range of user data, and a proxy class that emulates a variety of thermal and fast splits to check the eventual computed yield libraries.
2019-08-13 16:38:24 -05:00
Improve testing for FY helpers using C API 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.
2019-08-21 17:20:19 -05:00
actual_yields = helper.weighted_yields(0)
Add FissionYieldCutoffHelper for weighting yields by a cutoff energy This helper performs the following tasks to weight fission yields: 1) Determine a set of "fast" and "thermal" yields based on a user specified cutoff energy, 2) Tally fission rates above and below the cutoff, and 3) Compute the effective yield by weighting the fast and thermal yields by the fraction of fast and thermal fissions for all nuclides in all burnable materials. The user is allowed to specify the cutoff energy and energies preferred for the fast and thermal yields. By default, the cutoff is 112 eV, chosen as it is the logarithmic mean of 0.0253 eV and 500 keV. These two values are the default energies for thermal and fast yields. Tests are added to check for failure in constructing this helper (thermal energy > cutoff, non-real values passed, etc.), selection of yields given a range of user data, and a proxy class that emulates a variety of thermal and fast splits to check the eventual computed yield libraries.
2019-08-13 16:38:24 -05:00
assert actual_yields["U238"] == nuclide_bundle.u238.yield_data[5e5]
Improve testing for FY helpers using C API 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.
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for nuc in tally_nucs:
assert actual_yields[nuc] == (
helper.thermal_yields[nuc] * therm_frac
+ helper.fast_yields[nuc] * (1 - therm_frac))
Add AveragedFissionYieldHelper operator helper Computes the effective fission yields for nuclides with multiple sets of yields by 1) Tallying fission rate and energy-weighted fission rate 2) Determining average energy at which fission events occur 3) Interpolating [lin-lin] between adjacent fission yields based on this average energy Uses a second tally with an EnergyFunctionFilter to tally E * sigma_f * phi. Added unit tests with the proxy-style class that doesn't use the C API to generate tallies but tests all other aspects of the implementation. Tests cover three possible average energies (below minimum supplied yield energy, above max supplied yield energy, and between two) The Operator uses fission_yield_mode="average" to select this helper. The helper also has a from_operator class method.
2019-08-14 17:39:57 -05:00
Improve testing for FY helpers using C API 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.
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@pytest.mark.parametrize("avg_energy", (0.01, 6e5, 15e6))
def test_averaged_helper(materials, nuclide_bundle, avg_energy):
helper = AveragedFissionYieldHelper(nuclide_bundle)
helper.generate_tallies(materials, [0])
tallied_nucs = helper.update_tally_nuclides(
[n.name for n in nuclide_bundle])
Return list of str from FY helpers update_tally_nuclides
2019-08-22 15:32:45 -05:00
assert tallied_nucs == ["Pu239", "U235"]
Add AveragedFissionYieldHelper operator helper Computes the effective fission yields for nuclides with multiple sets of yields by 1) Tallying fission rate and energy-weighted fission rate 2) Determining average energy at which fission events occur 3) Interpolating [lin-lin] between adjacent fission yields based on this average energy Uses a second tally with an EnergyFunctionFilter to tally E * sigma_f * phi. Added unit tests with the proxy-style class that doesn't use the C API to generate tallies but tests all other aspects of the implementation. Tests cover three possible average energies (below minimum supplied yield energy, above max supplied yield energy, and between two) The Operator uses fission_yield_mode="average" to select this helper. The helper also has a from_operator class method.
2019-08-14 17:39:57 -05:00
Improve testing for FY helpers using C API 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.
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# check generated tallies
fission_tally = helper._fission_rate_tally
assert fission_tally is not None
fission_filters = fission_tally.filters
assert len(fission_filters) == 2
Rename openmc.capi -> openmc.lib
2019-09-05 07:31:13 -05:00
assert isinstance(fission_filters[0], lib.MaterialFilter)
Improve testing for FY helpers using C API 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.
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assert len(fission_filters[0].bins) == len(materials)
Rename openmc.capi -> openmc.lib
2019-09-05 07:31:13 -05:00
assert isinstance(fission_filters[1], lib.EnergyFilter)
Improve testing for FY helpers using C API 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.
2019-08-21 17:20:19 -05:00
assert len(fission_filters[1].bins) == 2
assert fission_tally.scores == ["fission"]
assert fission_tally.nuclides == list(tallied_nucs)
Add AveragedFissionYieldHelper operator helper Computes the effective fission yields for nuclides with multiple sets of yields by 1) Tallying fission rate and energy-weighted fission rate 2) Determining average energy at which fission events occur 3) Interpolating [lin-lin] between adjacent fission yields based on this average energy Uses a second tally with an EnergyFunctionFilter to tally E * sigma_f * phi. Added unit tests with the proxy-style class that doesn't use the C API to generate tallies but tests all other aspects of the implementation. Tests cover three possible average energies (below minimum supplied yield energy, above max supplied yield energy, and between two) The Operator uses fission_yield_mode="average" to select this helper. The helper also has a from_operator class method.
2019-08-14 17:39:57 -05:00
Improve testing for FY helpers using C API 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.
2019-08-21 17:20:19 -05:00
weighted_tally = helper._weighted_tally
assert weighted_tally is not None
weighted_filters = weighted_tally.filters
assert len(weighted_filters) == 2
Rename openmc.capi -> openmc.lib
2019-09-05 07:31:13 -05:00
assert isinstance(weighted_filters[0], lib.MaterialFilter)
Improve testing for FY helpers using C API 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.
2019-08-21 17:20:19 -05:00
assert len(weighted_filters[0].bins) == len(materials)
Rename openmc.capi -> openmc.lib
2019-09-05 07:31:13 -05:00
assert isinstance(weighted_filters[1], lib.EnergyFunctionFilter)
Improve testing for FY helpers using C API 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.
2019-08-21 17:20:19 -05:00
assert len(weighted_filters[1].energy) == 2
assert len(weighted_filters[1].y) == 2
assert weighted_tally.scores == ["fission"]
assert weighted_tally.nuclides == list(tallied_nucs)
helper_flux = 1e16
fission_results = proxy_tally_data(fission_tally, helper_flux)
weighted_results = proxy_tally_data(
weighted_tally, helper_flux * avg_energy)
helper._fission_rate_tally = Mock()
helper._weighted_tally = Mock()
Use tally mean in helpers rather than direct access to results
2020-07-22 22:50:05 -05:00
helper._fission_rate_tally.mean = fission_results
helper._weighted_tally.mean = weighted_results
Improve testing for FY helpers using C API 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.
2019-08-21 17:20:19 -05:00
helper.unpack()
Use python API for setup in test_deplete_fission_yields
2019-08-30 09:55:58 -05:00
expected_results = np.ones((1, len(tallied_nucs))) * avg_energy
Improve testing for FY helpers using C API 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.
2019-08-21 17:20:19 -05:00
assert helper.results == pytest.approx(expected_results)
actual_yields = helper.weighted_yields(0)
Add AveragedFissionYieldHelper operator helper Computes the effective fission yields for nuclides with multiple sets of yields by 1) Tallying fission rate and energy-weighted fission rate 2) Determining average energy at which fission events occur 3) Interpolating [lin-lin] between adjacent fission yields based on this average energy Uses a second tally with an EnergyFunctionFilter to tally E * sigma_f * phi. Added unit tests with the proxy-style class that doesn't use the C API to generate tallies but tests all other aspects of the implementation. Tests cover three possible average energies (below minimum supplied yield energy, above max supplied yield energy, and between two) The Operator uses fission_yield_mode="average" to select this helper. The helper also has a from_operator class method.
2019-08-14 17:39:57 -05:00
# constant U238 => no interpolation
assert actual_yields["U238"] == nuclide_bundle.u238.yield_data[5e5]
# construct expected yields
Improve testing for FY helpers using C API 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.
2019-08-21 17:20:19 -05:00
exp_u235_yields = interp_average_yields(nuclide_bundle.u235, avg_energy)
Add AveragedFissionYieldHelper operator helper Computes the effective fission yields for nuclides with multiple sets of yields by 1) Tallying fission rate and energy-weighted fission rate 2) Determining average energy at which fission events occur 3) Interpolating [lin-lin] between adjacent fission yields based on this average energy Uses a second tally with an EnergyFunctionFilter to tally E * sigma_f * phi. Added unit tests with the proxy-style class that doesn't use the C API to generate tallies but tests all other aspects of the implementation. Tests cover three possible average energies (below minimum supplied yield energy, above max supplied yield energy, and between two) The Operator uses fission_yield_mode="average" to select this helper. The helper also has a from_operator class method.
2019-08-14 17:39:57 -05:00
assert actual_yields["U235"] == exp_u235_yields
Improve testing for FY helpers using C API 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.
2019-08-21 17:20:19 -05:00
exp_pu239_yields = interp_average_yields(nuclide_bundle.pu239, avg_energy)
assert actual_yields["Pu239"] == exp_pu239_yields
def interp_average_yields(nuc, avg_energy):
"""Construct a set of yields by interpolation between neighbors"""
energies = nuc.yield_energies
yields = nuc.yield_data
if avg_energy < energies[0]:
return yields[energies[0]]
if avg_energy > energies[-1]:
return yields[energies[-1]]
thermal_ix = bisect.bisect_left(energies, avg_energy)
thermal_E, fast_E = energies[thermal_ix - 1:thermal_ix + 1]
assert thermal_E < avg_energy < fast_E
split = (avg_energy - thermal_E)/(fast_E - thermal_E)
return yields[thermal_E]*(1 - split) + yields[fast_E]*split
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