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Fixed a merge conflict issue.

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This commit is contained in:
Adam Nelson 2014-06-27 06:51:37 -04:00
parent 2153223272
commit c9561a8884
1 changed files with 10 additions and 10 deletions
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20
src/tally.F90
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@ -4,7 +4,7 @@ module tally
use constants
use error, only: fatal_error
use global
use math, only: t_percentile, calc_pn
use math, only: t_percentile, calc_pn, calc_rn
use mesh, only: get_mesh_bin, bin_to_mesh_indices, &
get_mesh_indices, mesh_indices_to_bin, &
mesh_intersects_2d, mesh_intersects_3d
@ -244,7 +244,7 @@ contains
score = last_wgt
case (SCORE_NU_SCATTER)
case (SCORE_NU_SCATTER)
! Skip any event where the particle didn't scatter
if (p % event /= EVENT_SCATTER) cycle SCORE_LOOP
@ -253,7 +253,7 @@ contains
! reaction with neutrons in the exit channel
score = wgt
case (SCORE_SCATTER_N)
! Skip any event where the particle didn't scatter
if (p % event /= EVENT_SCATTER) cycle SCORE_LOOP
@ -281,7 +281,7 @@ contains
score_index = score_index + 1
! get the score and tally it
score = last_wgt * calc_pn(n, mu)
!$omp critical
t % results(score_index, filter_index) % value = &
t % results(score_index, filter_index) % value + score
@ -503,7 +503,7 @@ contains
end if
end if
case (SCORE_KAPPA_FISSION)
if (survival_biasing) then
! No fission events occur if survival biasing is on -- need to
@ -977,7 +977,7 @@ contains
do l = 1, mat % n_nuclides
! Get atom density
atom_density = mat % atom_density(l)
! Get index in nuclides array
i_nuc = mat % nuclide(l)
@ -1179,7 +1179,7 @@ contains
! sections that are used often (e.g. n2n, ngamma, etc. for depletion),
! it might make sense to optimize this section or pre-calculate cross
! sections
if (score_bin > 1) then
! Set default score
score = ZERO
@ -1318,7 +1318,7 @@ contains
! Any other cross section has to be calculated on-the-fly. This is
! somewhat costly since it requires a loop over each nuclide in a
! material and each reaction in the nuclide
if (score_bin > 1) then
! Set default score
score = ZERO
@ -1630,7 +1630,7 @@ contains
score_index = (b - 1)*t % n_score_bins + j
if (i_nuclide > 0) then
! Determine macroscopic nuclide cross section
! Determine macroscopic nuclide cross section
select case(score_bin)
case (SCORE_FLUX)
score = flux
@ -1713,7 +1713,7 @@ contains
end select
else
! Determine macroscopic material cross section
! Determine macroscopic material cross section
select case(score_bin)
case (SCORE_FLUX)
score = flux