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heating analog tally for specific nuclide

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This commit is contained in:
liangjg 2019-04-06 08:02:07 -04:00
parent 5db6db53f1
commit 63527fe908
1 changed files with 49 additions and 23 deletions
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72
src/tallies/tally_scoring.cpp
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@ -1153,28 +1153,9 @@ score_general_ce(Particle* p, int i_tally, int start_index,
score = 0.;
if (p->type_ == Particle::Type::neutron) {
if (tally.estimator_ == ESTIMATOR_ANALOG) {
// Calculate material heating cross section
double macro_heating = 0.;
const Material& material {*model::materials[p->material_]};
for (auto i = 0; i < material.nuclide_.size(); ++i) {
auto j_nuclide = material.nuclide_[i];
auto atom_density = material.atom_density_(i);
const auto& nuc {*data::nuclides[j_nuclide]};
auto m = nuc.reaction_index_[NEUTRON_HEATING];
if (m == C_NONE) continue;
const auto& rxn {*nuc.reactions_[m]};
auto i_temp = p->neutron_xs_[j_nuclide].index_temp;
if (i_temp >= 0) { // Can be false due to multipole
auto i_grid = p->neutron_xs_[j_nuclide].index_grid;
auto f = p->neutron_xs_[j_nuclide].interp_factor;
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;
}
}
}
// All events score to an heating rate bin
// All events score to a heating tally bin. We actually use a
// collision estimator in place of an analog one since there is no
// reaction-wise heating cross section
if (settings::survival_biasing) {
// We need to account for the fact that some weight was already
// absorbed
@ -1182,7 +1163,52 @@ score_general_ce(Particle* p, int i_tally, int start_index,
} else {
score = p->wgt_last_;
}
score *= macro_heating * flux / p->macro_xs_.total;
if (i_nuclide >= 0) {
// Calculate nuclide heating cross section
double macro_heating = 0.;
const auto& nuc {*data::nuclides[i_nuclide]};
auto m = nuc.reaction_index_[NEUTRON_HEATING];
if (m == C_NONE) continue;
const auto& rxn {*nuc.reactions_[m]};
auto i_temp = p->neutron_xs_[i_nuclide].index_temp;
if (i_temp >= 0) { // Can be false due to multipole
auto i_grid = p->neutron_xs_[i_nuclide].index_grid;
auto f = p->neutron_xs_[i_nuclide].interp_factor;
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;
}
}
score *= macro_heating * flux / p->neutron_xs_[i_nuclide].total;
} else {
if (p->material_ != MATERIAL_VOID) {
// Calculate material heating cross section
double macro_heating = 0.;
const Material& material {*model::materials[p->material_]};
for (auto i = 0; i < material.nuclide_.size(); ++i) {
auto j_nuclide = material.nuclide_[i];
auto atom_density = material.atom_density_(i);
const auto& nuc {*data::nuclides[j_nuclide]};
auto m = nuc.reaction_index_[NEUTRON_HEATING];
if (m == C_NONE) continue;
const auto& rxn {*nuc.reactions_[m]};
auto i_temp = p->neutron_xs_[j_nuclide].index_temp;
if (i_temp >= 0) { // Can be false due to multipole
auto i_grid = p->neutron_xs_[j_nuclide].index_grid;
auto f = p->neutron_xs_[j_nuclide].interp_factor;
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;
}
}
}
score *= macro_heating * flux / p->macro_xs_.total;
} else {
score = 0.;
}
}
} else {
// Calculate neutron heating cross section on-the-fly
if (i_nuclide >= 0) {