From 1e315e7af3d720aa06c0c07050e0e5bf4667e7e5 Mon Sep 17 00:00:00 2001 From: Paul Romano Date: Thu, 13 Apr 2017 09:32:30 -0500 Subject: [PATCH] Fix adding fission energy release --- openmc/data/neutron.py | 4 ++-- 1 file changed, 2 insertions(+), 2 deletions(-) diff --git a/openmc/data/neutron.py b/openmc/data/neutron.py index 4da0d08c38..1d3a0176d4 100644 --- a/openmc/data/neutron.py +++ b/openmc/data/neutron.py @@ -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)