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Make sure heating is calculated correctly with threshold fission

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When trying to process Pa233 from ENDF/B-VIII.0, which has a threshold fission
reaction, I discovered that calculating heating numbers crashed because we were
trying to do arithmetic on arrays of different sizes. Re-evaluting the fission
cross section on the proper energy grid does the trick here.
This commit is contained in:
Paul Romano 2019-10-02 21:09:18 -05:00
parent 87541e969c
commit e510064b9c
1 changed files with 2 additions and 2 deletions
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4
openmc/data/neutron.py
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@ -874,7 +874,7 @@ class IncidentNeutron(EqualityMixin):
fission = data.reactions[18].xs[temp]
kerma_fission = get_file3_xs(ev, 318, E)
kerma.y = kerma.y - kerma_fission + (
f.fragments(E) + f.betas(E)) * fission.y
f.fragments(E) + f.betas(E)) * fission(E)
# For local KERMA, we first need to get the values from the
# HEATR run with photon energy deposited locally and put
@ -887,7 +887,7 @@ class IncidentNeutron(EqualityMixin):
kerma_fission_local = get_file3_xs(ev_local, 318, E)
kerma_local = kerma_local - kerma_fission_local + (
f.fragments(E) + f.prompt_photons(E)
+ f.delayed_photons(E) + f.betas(E))*fission.y
+ f.delayed_photons(E) + f.betas(E))*fission(E)
heating_local.xs[temp] = Tabulated1D(E, kerma_local)