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Revert to reaction_index name

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This commit is contained in:
Paul Romano 2017-12-22 21:17:38 +07:00
parent 12ce15978d
commit 26eedca662
3 changed files with 9 additions and 10 deletions
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2
src/cross_section.F90
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@ -333,7 +333,7 @@ contains
! Depletion-related reactions
if (in_active) then
do j = 1, 6
i_rxn = nuc % rxn_index_MT(DEPLETION_RX(j))
i_rxn = nuc % reaction_index(DEPLETION_RX(j))
if (i_rxn > 0) then
associate (xs => nuc % reactions(i_rxn) % xs(i_temp))
if (i_grid >= xs % threshold) then

7
src/nuclide_header.F90
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@ -90,7 +90,7 @@ module nuclide_header
! Array that maps MT values to index in reactions; used at tally-time. Note
! that ENDF-102 does not assign any MT values above 891.
integer :: rxn_index_MT(891)
integer :: reaction_index(891)
! Fission energy release
class(Function1D), allocatable :: fission_q_prompt ! prompt neutrons, gammas
@ -575,7 +575,7 @@ contains
n_temperature = size(this % kTs)
allocate(this % sum_xs(n_temperature))
this % rxn_index_MT(:) = 0
this % reaction_index(:) = 0
do i = 1, n_temperature
! Allocate and initialize derived cross sections
n_grid = size(this % grid(i) % energy)
@ -594,9 +594,8 @@ contains
i_fission = 0
do i = 1, size(this % reactions)
call MTs % push_back(this % reactions(i) % MT)
this % rxn_index_MT(this % reactions(i) % MT) = i
this % reaction_index(this % reactions(i) % MT) = i
associate (rx => this % reactions(i))
! Skip total inelastic level scattering, gas production cross sections

10
src/tallies/tally.F90
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@ -247,7 +247,7 @@ contains
! multiplicities of one.
score = p % last_wgt * flux
else
m = nuclides(p % event_nuclide) % rxn_index_MT(p % event_MT)
m = nuclides(p % event_nuclide) % reaction_index(p % event_MT)
! Get yield and apply to score
associate (rxn => nuclides(p % event_nuclide) % reactions(m))
@ -273,7 +273,7 @@ contains
! multiplicities of one.
score = p % last_wgt * flux
else
m = nuclides(p % event_nuclide) % rxn_index_MT(p % event_MT)
m = nuclides(p % event_nuclide) % reaction_index(p % event_MT)
! Get yield and apply to score
associate (rxn => nuclides(p % event_nuclide) % reactions(m))
@ -299,7 +299,7 @@ contains
! multiplicities of one.
score = p % last_wgt * flux
else
m = nuclides(p % event_nuclide) % rxn_index_MT(p % event_MT)
m = nuclides(p % event_nuclide) % reaction_index(p % event_MT)
! Get yield and apply to score
associate (rxn => nuclides(p%event_nuclide)%reactions(m))
@ -1232,7 +1232,7 @@ contains
score = ZERO
if (i_nuclide > 0) then
m = nuclides(i_nuclide) % rxn_index_MT(score_bin)
m = nuclides(i_nuclide) % reaction_index(score_bin)
if (m /= 0) then
! Retrieve temperature and energy grid index and interpolation
! factor
@ -1265,7 +1265,7 @@ contains
! Get index in nuclides array
i_nuc = materials(p % material) % nuclide(l)
m = nuclides(i_nuc) % rxn_index_MT(score_bin)
m = nuclides(i_nuc) % reaction_index(score_bin)
if (m /= 0) then
! Retrieve temperature and energy grid index and
! interpolation factor