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Find nu properly for nuclides with first-, second-, etc. chance fission

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This commit is contained in:
Paul Romano 2018-11-13 10:49:04 -06:00
parent 1b07521704
commit 927cbdac14
1 changed files with 2 additions and 2 deletions
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4
openmc/data/fission_energy.py
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@ -279,11 +279,11 @@ class FissionEnergyRelease(EqualityMixin):
# the delayed neutron fraction is so small that the difference
# is negligible. MT=18 (n, fission) might not be available so
# try MT=19 (n, f) as well.
if 18 in incident_neutron.reactions:
if 18 in incident_neutron and not incident_neutron[18].redundant:
nu = [p.yield_ for p in incident_neutron[18].products
if p.particle == 'neutron'
and p.emission_mode in ('prompt', 'total')]
elif 19 in incident_neutron.reactions:
elif 19 in incident_neutron:
nu = [p.yield_ for p in incident_neutron[19].products
if p.particle == 'neutron'
and p.emission_mode in ('prompt', 'total')]