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fix tallying of nculide-specific heating for neutrons

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This commit is contained in:
liangjg 2019-04-10 16:18:26 -04:00
parent 7536de5644
commit e7ab0b704c
1 changed files with 4 additions and 5 deletions
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9
src/tallies/tally_scoring.cpp
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@ -1177,7 +1177,7 @@ score_general_ce(Particle* p, int i_tally, int start_index,
const auto& xs {rxn.xs_[i_temp]};
if (i_grid >= xs.threshold) {
macro_heating = ((1.0 - f) * xs.value[i_grid-xs.threshold]
+ f * xs.value[i_grid-xs.threshold+1]) * atom_density;
+ f * xs.value[i_grid-xs.threshold+1]);
}
}
score *= macro_heating * flux / p->neutron_xs_[i_nuclide].total;
@ -2130,12 +2130,11 @@ void score_analog_tally_ce(Particle* p)
auto i_nuclide = tally.nuclides_[i];
// Tally this event in the present nuclide bin if that bin represents
// the event nuclide or the total material. Note that the i_nuclide
// and flux arguments for score_general are not used for analog
// tallies.
// the event nuclide or the total material. Note that the atomic
// density argument for score_general is not used for analog tallies.
if (i_nuclide == p->event_nuclide_ || i_nuclide == -1)
score_general_ce(p, i_tally, i*tally.scores_.size(), filter_index,
-1, -1., filter_weight);
i_nuclide, -1., filter_weight);
}
} else {