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Added general scoring to score_all_nuclides
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Sterling Harper
parent
347610a153
commit
f24e14fbd5
1 changed files with 9 additions and 289 deletions
298
src/tally.F90
298
src/tally.F90
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@ -801,22 +801,11 @@ contains
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integer, intent(in) :: filter_index
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integer :: i ! loop index for nuclides in material
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integer :: j ! loop index for scoring bin types
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integer :: m ! loop index for reactions in nuclide
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integer :: n ! loop index for legendre order
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integer :: num_nm ! Number of N,M orders in harmonic
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integer :: q ! loop index for scoring bins
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integer :: i_nuclide ! index in nuclides array
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integer :: score_bin ! type of score, e.g. SCORE_FLUX
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integer :: score_index ! scoring bin index
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integer :: i_energy ! index in nuclide energy grid
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real(8) :: f ! interpolation factor
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real(8) :: score ! actual scoring tally value
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real(8) :: atom_density ! atom density of single nuclide in atom/b-cm
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type(TallyObject), pointer, save :: t => null()
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type(Material), pointer, save :: mat => null()
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type(Reaction), pointer, save :: rxn => null()
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!$omp threadprivate(t, mat, rxn)
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!$omp threadprivate(t, mat)
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! Get pointer to tally
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t => tallies(i_tally)
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@ -835,290 +824,21 @@ contains
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i_nuclide = mat % nuclide(i)
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atom_density = mat % atom_density(i)
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! Loop over score types for each bin
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j = 0
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SCORE_LOOP: do q = 1, t % n_user_score_bins
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j = j + 1
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! determine what type of score bin
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score_bin = t % score_bins(j)
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! Determine scoring bin index based on what the index of the nuclide
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! is in the nuclides array
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score_index = (i_nuclide - 1)*t % n_score_bins + j
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! Determine macroscopic nuclide cross section
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select case(score_bin)
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case (SCORE_FLUX)
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score = flux
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case (SCORE_FLUX_YN)
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score_index = score_index - 1
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! For flux, we need no cross section
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score = flux
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num_nm = 1
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! Find the order for a collection of requested moments
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! and store the moment contribution of each
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do n = 0, t % moment_order(j)
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! determine scoring bin index
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score_index = score_index + num_nm
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! Update number of total n,m bins for this n (m = [-n: n])
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num_nm = 2 * n + 1
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! multiply score by the angular flux moments and store
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!$omp critical
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t % results(score_index: score_index + num_nm - 1, filter_index) % value = &
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t % results(score_index: score_index + num_nm - 1, filter_index) % value + &
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score * calc_rn(n, p % coord0 % uvw)
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!$omp end critical
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end do
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j = j + (t % moment_order(j) + 1)**2 - 1
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cycle SCORE_LOOP
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case (SCORE_TOTAL)
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score = micro_xs(i_nuclide) % total * atom_density * flux
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case (SCORE_TOTAL_YN)
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score_index = score_index - 1
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score = micro_xs(i_nuclide) % total * atom_density * flux
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num_nm = 1
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! Find the order for a collection of requested moments
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! and store the moment contribution of each
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do n = 0, t % moment_order(j)
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! determine scoring bin index
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score_index = score_index + num_nm
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! Update number of total n,m bins for this n (m = [-n: n])
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num_nm = 2 * n + 1
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! multiply score by the angular flux moments and store
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!$omp critical
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t % results(score_index: score_index + num_nm - 1, filter_index) % value = &
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t % results(score_index: score_index + num_nm - 1, filter_index) % value + &
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score * calc_rn(n, p % coord0 % uvw)
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!$omp end critical
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end do
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j = j + (t % moment_order(j) + 1)**2 - 1
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cycle SCORE_LOOP
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case (SCORE_SCATTER)
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score = (micro_xs(i_nuclide) % total - &
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micro_xs(i_nuclide) % absorption) * atom_density * flux
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case (SCORE_ABSORPTION)
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score = micro_xs(i_nuclide) % absorption * atom_density * flux
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case (SCORE_FISSION)
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score = micro_xs(i_nuclide) % fission * atom_density * flux
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case (SCORE_NU_FISSION)
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score = micro_xs(i_nuclide) % nu_fission * atom_density * flux
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case (SCORE_KAPPA_FISSION)
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score = micro_xs(i_nuclide) % kappa_fission * atom_density * flux
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case (SCORE_EVENTS)
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score = ONE
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case default
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! Any other cross section has to be calculated on-the-fly. For cross
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! sections that are used often (e.g. n2n, ngamma, etc. for depletion),
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! it might make sense to optimize this section or pre-calculate cross
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! sections
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if (score_bin > 1) then
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! Set default score
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score = ZERO
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! TODO: The following search for the matching reaction could be
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! replaced by adding a dictionary on each Nuclide instance of the
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! form {MT: i_reaction, ...}
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REACTION_LOOP: do m = 1, nuclides(i_nuclide) % n_reaction
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! Get pointer to reaction
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rxn => nuclides(i_nuclide) % reactions(m)
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! Check if this is the desired MT
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if (score_bin == rxn % MT) then
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! Retrieve index on nuclide energy grid and interpolation factor
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i_energy = micro_xs(i_nuclide) % index_grid
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f = micro_xs(i_nuclide) % interp_factor
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if (i_energy >= rxn % threshold) then
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score = ((ONE - f) * rxn % sigma(i_energy - &
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rxn%threshold + 1) + f * rxn % sigma(i_energy - &
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rxn%threshold + 2)) * atom_density * flux
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end if
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exit REACTION_LOOP
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end if
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end do REACTION_LOOP
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else
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call fatal_error("Invalid score type on tally " &
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&// to_str(t % id) // ".")
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end if
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end select
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! Add score to tally
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!$omp atomic
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t % results(score_index, filter_index) % value = &
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t % results(score_index, filter_index) % value + score
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end do SCORE_LOOP
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! Determine score for each bin
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call score_general(p, t, (i_nuclide-1)*t % n_score_bins, filter_index, &
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&i_nuclide, atom_density, flux)
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end do NUCLIDE_LOOP
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! ==========================================================================
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! SCORE TOTAL MATERIAL REACTION RATES
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! Loop over score types for each bin
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j = 0
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MATERIAL_SCORE_LOOP: do q = 1, t % n_user_score_bins
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j = j + 1
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! determine what type of score bin
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score_bin = t % score_bins(j)
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i_nuclide = -1
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atom_density = ZERO
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! Determine scoring bin index based on what the index of the nuclide
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! is in the nuclides array
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score_index = n_nuclides_total*t % n_score_bins + j
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! Determine macroscopic material cross section
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select case(score_bin)
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case (SCORE_FLUX)
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score = flux
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case (SCORE_FLUX_YN)
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score_index = score_index - 1
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! For flux, we need no cross section
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score = flux
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num_nm = 1
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! Find the order for a collection of requested moments
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! and store the moment contribution of each
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do n = 0, t % moment_order(j)
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! determine scoring bin index
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score_index = score_index + num_nm
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! Update number of total n,m bins for this n (m = [-n: n])
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num_nm = 2 * n + 1
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! multiply score by the angular flux moments and store
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!$omp critical
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t % results(score_index: score_index + num_nm - 1, filter_index) % value = &
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t % results(score_index: score_index + num_nm - 1, filter_index) % value + &
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score * calc_rn(n, p % coord0 % uvw)
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!$omp end critical
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end do
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j = j + (t % moment_order(j) + 1)**2 - 1
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cycle MATERIAL_SCORE_LOOP
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case (SCORE_TOTAL)
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score = material_xs % total * flux
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case (SCORE_TOTAL_YN)
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score_index = score_index - 1
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! Total cross section is pre-calculated
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score = material_xs % total * flux
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num_nm = 1
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! Find the order for a collection of requested moments
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! and store the moment contribution of each
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do n = 0, t % moment_order(j)
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! determine scoring bin index
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score_index = score_index + num_nm
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! Update number of total n,m bins for this n (m = [-n: n])
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num_nm = 2 * n + 1
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! multiply score by the angular flux moments and store
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!$omp critical
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t % results(score_index: score_index + num_nm - 1, filter_index) % value = &
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t % results(score_index: score_index + num_nm - 1, filter_index) % value + &
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score * calc_rn(n, p % coord0 % uvw)
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!$omp end critical
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end do
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j = j + (t % moment_order(j) + 1)**2 - 1
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cycle MATERIAL_SCORE_LOOP
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case (SCORE_SCATTER)
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score = (material_xs % total - material_xs % absorption) * flux
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case (SCORE_ABSORPTION)
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score = material_xs % absorption * flux
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case (SCORE_FISSION)
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score = material_xs % fission * flux
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case (SCORE_NU_FISSION)
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score = material_xs % nu_fission * flux
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case (SCORE_KAPPA_FISSION)
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score = material_xs % kappa_fission * flux
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case (SCORE_EVENTS)
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score = ONE
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case default
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! Any other cross section has to be calculated on-the-fly. This is
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! somewhat costly since it requires a loop over each nuclide in a
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! material and each reaction in the nuclide
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if (score_bin > 1) then
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! Set default score
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score = ZERO
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! Get pointer to current material
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mat => materials(p % material)
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do i = 1, mat % n_nuclides
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! Get atom density
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atom_density = mat % atom_density(i)
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! Get index in nuclides array
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i_nuclide = mat % nuclide(i)
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! TODO: The following search for the matching reaction could
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! be replaced by adding a dictionary on each Nuclide
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! instance of the form {MT: i_reaction, ...}
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do m = 1, nuclides(i_nuclide) % n_reaction
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! Get pointer to reaction
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rxn => nuclides(i_nuclide) % reactions(m)
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! Check if this is the desired MT
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if (score_bin == rxn % MT) then
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! Retrieve index on nuclide energy grid and interpolation
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! factor
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i_energy = micro_xs(i_nuclide) % index_grid
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f = micro_xs(i_nuclide) % interp_factor
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if (i_energy >= rxn % threshold) then
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score = score + ((ONE - f) * rxn % sigma(i_energy - &
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rxn%threshold + 1) + f * rxn % sigma(i_energy - &
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rxn%threshold + 2)) * atom_density * flux
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end if
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exit
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end if
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end do
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end do
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else
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call fatal_error("Invalid score type on tally " &
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&// to_str(t % id) // ".")
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end if
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end select
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! Add score to tally
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!$omp atomic
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t % results(score_index, filter_index) % value = &
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t % results(score_index, filter_index) % value + score
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end do MATERIAL_SCORE_LOOP
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! Determine score for each bin
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call score_general(p, t, n_nuclides_total*t % n_score_bins, filter_index, &
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&i_nuclide, atom_density, flux)
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end subroutine score_all_nuclides
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