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various small cleanups

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This commit is contained in:
John Tramm 2019-12-17 17:51:52 +00:00
parent c8755273a0
commit 44f0dd73e8
3 changed files with 5 additions and 92 deletions
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95
src/simulation.cpp
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@ -224,20 +224,11 @@ void process_calculate_xs_events(QueueItem * queue, int n)
if (p->write_track_) write_particle_track(*p);
if (settings::check_overlaps) check_cell_overlap(p);
if (settings::run_CE) {
if (p->material_ == p->material_last_ && p->sqrtkT_ != p->sqrtkT_last_) {
// Remove particle from queue
}
}
// Find energy index on energy grid
// TODO: Calculate this separately?
//int neutron = static_cast<int>(Particle::Type::neutron);
//p->macro_xs_.i_grid = std::log(p->E_/data::energy_min[neutron]) / simulation::log_spacing;
}
if( lost_particles > 0 )
exit(1);
#pragma omp parallel for
for( int i = 0; i < n; i++ )
{
@ -258,13 +249,8 @@ void process_calculate_xs_events(QueueItem * queue, int n)
p->macro_xs_.fission = 0.0;
p->macro_xs_.nu_fission = 0.0;
}
/*
int idx;
#pragma omp atomic capture
idx = advance_particle_queue_length++;
advance_particle_queue[idx] = queue[i];
*/
}
int start = advance_particle_queue_length;
int end = start + n;
int j = 0;
@ -276,81 +262,6 @@ void process_calculate_xs_events(QueueItem * queue, int n)
j++;
}
advance_particle_queue_length += n;
/*
// Calculate nuclide micros
for (int i = 0; i < data::nuclides.size(); ++i) {
// loop over particles
for (int j = 0; j < n; j++) {
//Particle * p = particles + queue[i];
Particle * p = particles + queue[j];
if (p->material_ == MATERIAL_VOID) continue;
// If material doesn't have this nuclide, skip it
const auto& mat {model::materials[p->material_]};
if (mat->mat_nuclide_index_[i] == -1) continue;
// ======================================================================
// CHECK FOR S(A,B) TABLE
// Check if this nuclide matches one of the S(a,b) tables specified.
// This relies on thermal_tables_ being sorted by .index_nuclide
int i_sab = C_NONE;
double sab_frac = 0.0;
for (const auto& sab : mat->thermal_tables_) {
if (i == sab.index_nuclide) {
// Get index in sab_tables
i_sab = sab.index_table;
sab_frac = sab.fraction;
// If particle energy is greater than the highest energy for the
// S(a,b) table, then don't use the S(a,b) table
//if (p->E_ > data::thermal_scatt[i_sab]->threshold()) i_sab = C_NONE;
std::cout << "S(alpha, beta) max energy = " << data::thermal_scatt[i_sab]->energy_max_ <<std::endl;
if (p->E_ > data::thermal_scatt[i_sab]->energy_max_) i_sab = C_NONE;
}
}
// ======================================================================
// CALCULATE MICROSCOPIC CROSS SECTION
// Calculate microscopic cross section for this nuclide
const auto& micro {p->neutron_xs_[i]};
if (p->E_ != micro.last_E
|| p->sqrtkT_ != micro.last_sqrtkT
|| i_sab != micro.index_sab
|| sab_frac != micro.sab_frac) {
data::nuclides[i]->calculate_xs(i_sab, p->macro_xs_.i_grid, sab_frac, *p);
}
}
}
for (int i = 0; i < n; i++) {
Particle * p = particles + queue[i];
// Calculate microscopic and macroscopic cross sections
if (p->material_ != MATERIAL_VOID) {
// Only works for CE, no MG support
//if (settings::run_CE) {
// If the material is the same as the last material and the
// temperature hasn't changed, we don't need to lookup cross
// sections again.
model::materials[p->material_]->calculate_xs(*p);
}
//else {
//}
else {
p->macro_xs_.total = 0.0;
p->macro_xs_.absorption = 0.0;
p->macro_xs_.fission = 0.0;
p->macro_xs_.nu_fission = 0.0;
}
int idx;
#pragma omp atomic capture
idx = advance_particle_queue_length++;
advance_particle_queue[idx] = queue[i];
}
*/
}
void process_advance_particle_events()