/**
* @brief Normalizes all FSR scalar fluxes and Track boundary angular
* fluxes to the total fission source (times \f$ \nu \f$).
*/
void VectorizedSolver::normalizeFluxes() {
FP_PRECISION* nu_sigma_f;
FP_PRECISION volume;
FP_PRECISION tot_fission_source;
FP_PRECISION norm_factor;
/* Compute total fission source for each FSR, energy group */
#pragma omp parallel for private(volume, nu_sigma_f) \
reduction(+:tot_fission_source) schedule(guided)
for (int r=0; r < _num_FSRs; r++) {
/* Get pointers to important data structures */
nu_sigma_f = _FSR_materials[r]->getNuSigmaF();
volume = _FSR_volumes[r];
/* Loop over energy group vector lengths */
for (int v=0; v < _num_vector_lengths; v++) {
/* Loop over each energy group within this vector */
#pragma simd vectorlength(VEC_LENGTH)
for (int e=v*VEC_LENGTH; e < (v+1)*VEC_LENGTH; e++) {
_fission_sources(r,e) = nu_sigma_f[e] * _scalar_flux(r,e);
_fission_sources(r,e) *= volume;
}
}
}
/* Compute the total fission source */
int size = _num_FSRs * _num_groups;
#ifdef SINGLE
tot_fission_source = cblas_sasum(size, _fission_sources, 1);
#else
tot_fission_source = cblas_dasum(size, _fission_sources, 1);
#endif
/* Compute the normalization factor */
norm_factor = 1.0 / tot_fission_source;
log_printf(DEBUG, "Tot. Fiss. Src. = %f, Normalization factor = %f",
tot_fission_source, norm_factor);
/* Normalize the FSR scalar fluxes */
#ifdef SINGLE
cblas_sscal(size, norm_factor, _scalar_flux, 1);
#else
cblas_dscal(size, norm_factor, _scalar_flux, 1);
#endif
/* Normalize the Track angular boundary fluxes */
size = 2 * _tot_num_tracks * _num_polar * _num_groups;
#ifdef SINGLE
cblas_sscal(size, norm_factor, _boundary_flux, 1);
#else
cblas_dscal(size, norm_factor, _boundary_flux, 1);
#endif
return;
}
/**
* @brief Normalizes all FSR scalar fluxes and Track boundary angular
* fluxes to the total fission source (times \f$ \nu \f$).
*/
void CPUSolver::normalizeFluxes() {
FP_PRECISION* nu_sigma_f;
FP_PRECISION volume;
FP_PRECISION tot_fission_source;
FP_PRECISION norm_factor;
/* Compute total fission source for each FSR, energy group */
#pragma omp parallel for private(volume, nu_sigma_f) \
reduction(+:tot_fission_source) schedule(guided)
for (int r=0; r < _num_FSRs; r++) {
/* Get pointers to important data structures */
nu_sigma_f = _FSR_materials[r]->getNuSigmaF();
volume = _FSR_volumes[r];
for (int e=0; e < _num_groups; e++)
_fission_sources(r,e) = nu_sigma_f[e] * _scalar_flux(r,e) * volume;
}
/* Compute the total fission source */
tot_fission_source = pairwise_sum<FP_PRECISION>(_fission_sources,
_num_FSRs*_num_groups);
/* Normalize scalar fluxes in each FSR */
norm_factor = 1.0 / tot_fission_source;
log_printf(DEBUG, "Tot. Fiss. Src = %f, Normalization factor = %f",
tot_fission_source, norm_factor);
#pragma omp parallel for schedule(guided)
for (int r=0; r < _num_FSRs; r++) {
for (int e=0; e < _num_groups; e++)
_scalar_flux(r,e) *= norm_factor;
}
/* Normalize angular boundary fluxes for each Track */
#pragma omp parallel for schedule(guided)
for (int i=0; i < _tot_num_tracks; i++) {
for (int j=0; j < 2; j++) {
for (int p=0; p < _num_polar; p++) {
for (int e=0; e < _num_groups; e++) {
_boundary_flux(i,j,p,e) *= norm_factor;
}
}
}
}
return;
}
/**
* @brief Computes the total source (fission and scattering) in each FSR.
* @details This method computes the total source in each FSR based on
* this iteration's current approximation to the scalar flux. A
* residual for the source with respect to the source compute on
* the previous iteration is computed and returned. The residual
* is determined as follows:
* /f$ res = \sqrt{\frac{\displaystyle\sum \displaystyle\sum
* \left(\frac{Q^i - Q^{i-1}{Q^i}\right)^2}{\# FSRs}}} \f$
*
* @return the residual between this source and the previous source
*/
FP_PRECISION CPUSolver::computeFSRSources() {
int tid;
Material* material;
FP_PRECISION scatter_source;
FP_PRECISION fission_source;
FP_PRECISION* nu_sigma_f;
FP_PRECISION* sigma_s;
FP_PRECISION* sigma_t;
FP_PRECISION* chi;
FP_PRECISION source_residual = 0.0;
FP_PRECISION inverse_k_eff = 1.0 / _k_eff;
/* For all FSRs, find the source */
#pragma omp parallel for private(tid, material, nu_sigma_f, chi, \
sigma_s, sigma_t, fission_source, scatter_source) schedule(guided)
for (int r=0; r < _num_FSRs; r++) {
tid = omp_get_thread_num();
material = _FSR_materials[r];
nu_sigma_f = material->getNuSigmaF();
chi = material->getChi();
sigma_s = material->getSigmaS();
sigma_t = material->getSigmaT();
/* Initialize the source residual to zero */
_source_residuals[r] = 0.;
/* Compute fission source for each group */
if (material->isFissionable()) {
for (int e=0; e < _num_groups; e++)
_fission_sources(r,e) = _scalar_flux(r,e) * nu_sigma_f[e];
fission_source = pairwise_sum<FP_PRECISION>(&_fission_sources(r,0),
_num_groups);
fission_source *= inverse_k_eff;
}
else
fission_source = 0.0;
/* Compute total scattering source for group G */
for (int G=0; G < _num_groups; G++) {
scatter_source = 0;
for (int g=0; g < _num_groups; g++)
_scatter_sources(tid,g) = sigma_s[G*_num_groups+g] * _scalar_flux(r,g);
scatter_source=pairwise_sum<FP_PRECISION>(&_scatter_sources(tid,0),
_num_groups);
/* Set the total source for FSR r in group G */
_source(r,G) = (fission_source * chi[G] + scatter_source) *
ONE_OVER_FOUR_PI;
_reduced_source(r,G) = _source(r,G) / sigma_t[G];
/* Compute the norm of residual of the source in the FSR */
if (fabs(_source(r,G)) > 1E-10)
_source_residuals[r] += pow((_source(r,G) - _old_source(r,G))
/ _source(r,G), 2);
/* Update the old source */
_old_source(r,G) = _source(r,G);
}
}
/* Sum up the residuals from each FSR */
source_residual = pairwise_sum<FP_PRECISION>(_source_residuals, _num_FSRs);
source_residual = sqrt(source_residual / (_num_FSRs * _num_groups));
return source_residual;
}
/**
* @brief Computes the total source (fission and scattering) in each FSR.
* @details This method computes the total source in each FSR based on
* this iteration's current approximation to the scalar flux. A
* residual for the source with respect to the source compute on
* the previous iteration is computed and returned. The residual
* is determined as follows:
* /f$ res = \sqrt{\frac{\displaystyle\sum \displaystyle\sum
* \left(\frac{Q^i - Q^{i-1}{Q^i}\right)^2}{# FSRs}}} \f$
*
* @return the residual between this source and the previous source
*/
FP_PRECISION VectorizedSolver::computeFSRSources() {
int tid;
FP_PRECISION scatter_source;
FP_PRECISION fission_source;
FP_PRECISION* nu_sigma_f;
FP_PRECISION* sigma_s;
FP_PRECISION* sigma_t;
FP_PRECISION* chi;
Material* material;
FP_PRECISION source_residual = 0.0;
FP_PRECISION inverse_k_eff = 1.0 / _k_eff;
/* For all FSRs, find the source */
#pragma omp parallel for private(material, nu_sigma_f, chi, \
sigma_s, sigma_t, fission_source, scatter_source) schedule(guided)
for (int r=0; r < _num_FSRs; r++) {
tid = omp_get_thread_num();
material = _FSR_materials[r];
nu_sigma_f = material->getNuSigmaF();
chi = material->getChi();
sigma_s = material->getSigmaS();
sigma_t = material->getSigmaT();
/* Initialize the source residual to zero */
_source_residuals[r] = 0.;
/* Compute fission source for each group */
if (material->isFissionable()) {
for (int v=0; v < _num_vector_lengths; v++) {
/* Compute fission source for each group */
#pragma simd vectorlength(VEC_LENGTH)
for (int e=v*VEC_LENGTH; e < (v+1)*VEC_LENGTH; e++)
_fission_sources(r,e) = _scalar_flux(r,e) * nu_sigma_f[e];
}
#ifdef SINGLE
fission_source = cblas_sasum(_num_groups, &_fission_sources(r,0),1);
#else
fission_source = cblas_dasum(_num_groups, &_fission_sources(r,0),1);
#endif
fission_source *= inverse_k_eff;
}
else
fission_source = 0.0;
/* Compute total scattering source for group G */
for (int G=0; G < _num_groups; G++) {
scatter_source = 0;
for (int v=0; v < _num_vector_lengths; v++) {
#pragma simd vectorlength(VEC_LENGTH)
for (int g=v*VEC_LENGTH; g < (v+1)*VEC_LENGTH; g++)
_scatter_sources(tid,g) = sigma_s[G*_num_groups+g] *
_scalar_flux(r,g);
}
#ifdef SINGLE
scatter_source=cblas_sasum(_num_groups,&_scatter_sources(tid,0),1);
#else
scatter_source=cblas_dasum(_num_groups,&_scatter_sources(tid,0),1);
#endif
/* Set the total source for FSR r in group G */
_source(r,G) = (fission_source * chi[G] + scatter_source)
* ONE_OVER_FOUR_PI;
_reduced_source(r,G) = _source(r,G) / sigma_t[G];
/* Compute the norm of residual of the source in the FSR */
if (fabs(_source(r,G)) > 1E-10)
_source_residuals[r] += pow((_source(r,G) - _old_source(r,G))
/ _source(r,G), 2);
/* Update the old source */
_old_source(r,G) = _source(r,G);
}
}
/* Sum up the residuals from each group and in each FSR */
#ifdef SINGLE
source_residual = cblas_sasum(_num_FSRs,_source_residuals,1);
#else
source_residual = cblas_dasum(_num_FSRs,_source_residuals,1);
#endif
source_residual = sqrt(source_residual / (_num_groups * _num_FSRs));
return source_residual;
}
