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Ensure reaction rate tallies work when need_depletion_rx is false

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This commit is contained in:
Paul Romano 2020-08-24 13:08:18 -05:00
parent 2466f1739e
commit 234e09345d
1 changed files with 14 additions and 4 deletions
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18
src/tallies/tally_scoring.cpp
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@ -477,12 +477,13 @@ double get_nuclide_xs(const Particle& p, int i_nuclide, int score_bin) {
auto m = nuc.reaction_index_[score_bin];
if (m == C_NONE) return 0.0;
const auto& rxn {*nuc.reactions_[m]};
const auto& micro {p.neutron_xs_[i_nuclide]};
auto i_temp = p.neutron_xs_[i_nuclide].index_temp;
auto i_temp = micro.index_temp;
if (i_temp >= 0) { // Can be false due to multipole
// Get index on energy grid and interpolation factor
auto i_grid = p.neutron_xs_[i_nuclide].index_grid;
auto f = p.neutron_xs_[i_nuclide].interp_factor;
auto i_grid = micro.index_grid;
auto f = micro.interp_factor;
// Calculate interpolated cross section
const auto& xs {rxn.xs_[i_temp]};
@ -499,7 +500,7 @@ double get_nuclide_xs(const Particle& p, int i_nuclide, int score_bin) {
double kerma_fission = nuc.fragments_ ?
((*nuc.fragments_)(p.E_last_) + (*nuc.betas_)(p.E_last_) +
(*nuc.prompt_photons_)(p.E_last_) + (*nuc.delayed_photons_)(p.E_last_))
* p.neutron_xs_[i_nuclide].fission : 0.0;
* micro.fission : 0.0;
// Determine non-fission kerma as difference
double kerma_non_fission = value - kerma_fission;
@ -511,6 +512,9 @@ double get_nuclide_xs(const Particle& p, int i_nuclide, int score_bin) {
value = simulation::keff * kerma_non_fission + kerma_fission;
}
return value;
} else {
// For multipole, calculate (n,gamma) from other reactions
return rxn.mt_ == N_GAMMA ? micro.absorption - micro.fission : 0.0;
}
return 0.0;
}
@ -1271,6 +1275,11 @@ score_general_ce(Particle& p, int i_tally, int start_index, int filter_index,
case N_GAMMA:
case N_P:
case N_A:
// This case block only works if cross sections for these reactions have
// been precalculated. When they are not, we revert to the default case,
// which looks up cross sections
if (!simulation::need_depletion_rx) goto default_case;
if (p.type_ != Type::neutron) continue;
if (tally.estimator_ == TallyEstimator::ANALOG) {
@ -1365,6 +1374,7 @@ score_general_ce(Particle& p, int i_tally, int start_index, int filter_index,
break;
default:
default_case:
// The default block is really only meant for redundant neutron reactions
// (e.g. 444, 901)