crp/OpenMC
1
0
Fork 0
mirror of https://github.com/openmc-dev/openmc.git synced 2026-07-26 13:15:39 -04:00
Code Issues Projects Releases Packages Wiki Activity Actions

Make rxn_index_MT allocated as part of Nuclide type

Browse source
This commit is contained in:
Paul Romano 2017-12-22 13:11:17 +07:00
parent 29a8737b25
commit fe19b9b6dd
1 changed files with 6 additions and 5 deletions
Download patch file Download diff file Expand all files Collapse all files

11
src/nuclide_header.F90
View file

@ -87,9 +87,11 @@ module nuclide_header
! Reactions
type(Reaction), allocatable :: reactions(:)
!type(DictIntInt) :: reaction_index ! map MT values to index in reactions
! array; used at tally-time
integer, allocatable :: rxn_index_MT(:)
! 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)
! Fission energy release
class(Function1D), allocatable :: fission_q_prompt ! prompt neutrons, gammas
class(Function1D), allocatable :: fission_q_recov ! neutrons, gammas, betas
@ -561,8 +563,7 @@ contains
n_temperature = size(this % kTs)
allocate(this % sum_xs(n_temperature))
allocate(this % rxn_index_MT(1000))
this % rxn_index_MT(:)=0
this % rxn_index_MT(:) = 0
do i = 1, n_temperature
! Allocate and initialize derived cross sections
n_grid = size(this % grid(i) % energy)