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Fix adding fission energy release

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This commit is contained in:
Paul Romano 2017-04-13 09:32:30 -05:00
parent 31e5a9e081
commit 1e315e7af3
1 changed files with 2 additions and 2 deletions
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4
openmc/data/neutron.py
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@ -862,8 +862,8 @@ class IncidentNeutron(EqualityMixin):
# Add fission energy release data
ev = Evaluation(filename)
if ev.target['fissionable']:
data.fission_energy = FissionEnergyRelease.from_endf(ev)
if (1, 458) in ev.section:
data.fission_energy = FissionEnergyRelease.from_endf(ev, data)
# Add 0K elastic scattering cross section
pendf = Evaluation(pendf_file)