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Add MT=458 data library and supporting methods

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This commit is contained in:
Sterling Harper 2016-08-02 14:10:32 -05:00
parent 725ef2d923
commit 84cfd2d6a9
5 changed files with 155 additions and 63 deletions
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17
scripts/openmc-ace-to-hdf5
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@ -25,6 +25,13 @@ follows the NNDC data convention (1000*Z + A + 300 + 100*m), or the MCNP data
convention (essentially the same as NNDC, except that the first metastable state
of Am242 is 95242 and the ground state is 95642).
The optional --fission_energy_release argument will accept an HDF5 file
containing a library of fission energy release (ENDF MF=1 MT=458) data. A
library built from ENDF/B-VII.1 data is released with OpenMC and can be found at
openmc/data/fission_Q_data_endb71.h5. This data is necessary for
'fission-q-prompt' and 'fission-q-recoverable' tallies, but is not needed
otherwise.
"""
class CustomFormatter(argparse.ArgumentDefaultsHelpFormatter,
@ -47,6 +54,8 @@ parser.add_argument('--xsdir', help='MCNP xsdir file that lists '
'ACE libraries')
parser.add_argument('--xsdata', help='Serpent xsdata file that lists '
'ACE libraries')
parser.add_argument('--fission_energy_release', help='HDF5 file containing '
'fission energy release data')
args = parser.parse_args()
if not os.path.isdir(args.destination):
@ -111,6 +120,14 @@ for filename in ace_libraries:
# Continuous-energy neutron data
neutron = openmc.data.IncidentNeutron.from_ace(
table, args.metastable)
# Fission energy release data, if available
if args.fission_energy_release is not None:
fer = openmc.data.FissionEnergyRelease.from_compact_hdf5(
args.fission_energy_release, neutron)
if fer is not None:
neutron.fission_energy = fer
print(neutron.name)
# Determine filename