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Respond to comments by @wbinventor and @nelsonag on #500

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Paul Romano 2015-11-20 19:45:55 -06:00
parent 0927f25300
commit 088e865244
6 changed files with 44 additions and 44 deletions
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18
src/ace.F90
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@ -733,13 +733,13 @@ contains
! sigma array is not allocated or stored for elastic scattering since it is
! already stored in nuc % elastic
associate (rxn => nuc % reactions(1))
rxn % MT = 2
rxn % Q_value = ZERO
rxn % multiplicity = 1
rxn % threshold = 1
rxn % scatter_in_cm = .true.
rxn % has_angle_dist = .false.
rxn % has_energy_dist = .false.
rxn%MT = 2
rxn%Q_value = ZERO
rxn%multiplicity = 1
rxn%threshold = 1
rxn%scatter_in_cm = .true.
rxn%has_angle_dist = .false.
rxn%has_energy_dist = .false.
end associate
! Add contribution of elastic scattering to total cross section
@ -1013,8 +1013,8 @@ contains
recursive subroutine get_energy_dist(edist, loc_law, delayed_n)
type(DistEnergy), intent(inout) :: edist ! energy distribution
integer, intent(in) :: loc_law ! locator for data
logical, intent(in), optional :: delayed_n ! is this for delayed neutrons?
integer, intent(in) :: loc_law ! locator for data
logical, intent(in), optional :: delayed_n ! is this for delayed neutrons?
integer :: LDIS ! location of all energy distributions
integer :: LNW ! location of next energy distribution if multiple

10
src/cross_section.F90
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@ -510,8 +510,8 @@ contains
pure function find_energy_index(mat, E) result(i)
type(Material), intent(in) :: mat ! pointer to current material
real(8), intent(in) :: E ! energy of particle
integer :: i ! energy grid index
real(8), intent(in) :: E ! energy of particle
integer :: i ! energy grid index
! if the energy is outside of energy grid range, set to first or last
! index. Otherwise, do a binary search through the union energy grid.
@ -531,9 +531,9 @@ contains
!===============================================================================
pure function elastic_xs_0K(E, nuc) result(xs_out)
real(8), intent(in) :: E ! trial energy
type(Nuclide), intent(in) :: nuc ! target nuclide at temperature
real(8) :: xs_out ! 0K xs at trial energy
real(8), intent(in) :: E ! trial energy
type(Nuclide), intent(in) :: nuc ! target nuclide at temperature
real(8) :: xs_out ! 0K xs at trial energy
integer :: i_grid ! index on nuclide energy grid
real(8) :: f ! interp factor on nuclide energy grid

10
src/fission.F90
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@ -17,8 +17,8 @@ contains
pure function nu_total(nuc, E) result(nu)
type(Nuclide), intent(in) :: nuc ! nuclide from which to find nu
real(8), intent(in) :: E ! energy of incoming neutron
real(8) :: nu ! number of total neutrons emitted per fission
real(8), intent(in) :: E ! energy of incoming neutron
real(8) :: nu ! number of total neutrons emitted per fission
integer :: i ! loop index
integer :: NC ! number of polynomial coefficients
@ -50,8 +50,8 @@ contains
pure function nu_prompt(nuc, E) result(nu)
type(Nuclide), intent(in) :: nuc ! nuclide from which to find nu
real(8), intent(in) :: E ! energy of incoming neutron
real(8) :: nu ! number of prompt neutrons emitted per fission
real(8), intent(in) :: E ! energy of incoming neutron
real(8) :: nu ! number of prompt neutrons emitted per fission
integer :: i ! loop index
integer :: NC ! number of polynomial coefficients
@ -87,7 +87,7 @@ contains
pure function nu_delayed(nuc, E) result(nu)
type(Nuclide), intent(in) :: nuc ! nuclide from which to find nu
real(8), intent(in) :: E ! energy of incoming neutron
real(8), intent(in) :: E ! energy of incoming neutron
real(8) :: nu ! number of delayed neutrons emitted per fission
if (nuc % nu_d_type == NU_NONE) then

17
src/geometry.F90
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@ -875,6 +875,7 @@ contains
integer :: i ! index in cells/surfaces array
integer :: j ! index in region specification
integer :: k ! surface half-space spec
integer :: n ! size of vector
type(VectorInt), allocatable :: neighbor_pos(:)
type(VectorInt), allocatable :: neighbor_neg(:)
@ -902,17 +903,17 @@ contains
do i = 1, n_surfaces
! Copy positive neighbors to Surface instance
j = neighbor_pos(i)%size()
if (j > 0) then
allocate(surfaces(i)%obj%neighbor_pos(j))
surfaces(i)%obj%neighbor_pos(:) = neighbor_pos(i)%data(1:j)
n = neighbor_pos(i)%size()
if (n > 0) then
allocate(surfaces(i)%obj%neighbor_pos(n))
surfaces(i)%obj%neighbor_pos(:) = neighbor_pos(i)%data(1:n)
end if
! Copy negative neighbors to Surface instance
j = neighbor_neg(i)%size()
if (j > 0) then
allocate(surfaces(i)%obj%neighbor_neg(j))
surfaces(i)%obj%neighbor_neg(:) = neighbor_neg(i)%data(1:j)
n = neighbor_neg(i)%size()
if (n > 0) then
allocate(surfaces(i)%obj%neighbor_neg(n))
surfaces(i)%obj%neighbor_neg(:) = neighbor_neg(i)%data(1:n)
end if
end do

9
src/output.F90
View file

@ -194,8 +194,8 @@ contains
!===============================================================================
subroutine write_message(message, level)
character(*), intent(in) :: message
integer, intent(in), optional :: level ! verbosity level
character(*), intent(in) :: message ! message to write
integer, intent(in), optional :: level ! verbosity level
integer :: i_start ! starting position
integer :: i_end ! ending position
@ -1349,10 +1349,9 @@ contains
!===============================================================================
function get_label(t, i_filter) result(label)
type(TallyObject), intent(in) :: t ! tally object
integer, intent(in) :: i_filter ! index in filters array
character(100) :: label ! user-specified identifier
integer, intent(in) :: i_filter ! index in filters array
character(100) :: label ! user-specified identifier
integer :: i ! index in cells/surfaces/etc array
integer :: bin

24
src/tally.F90
View file

@ -608,13 +608,13 @@ contains
! No fission events occur if survival biasing is on -- need to
! calculate fraction of absorptions that would have resulted in
! fission scale by kappa-fission
associate (nuc => nuclides(p % event_nuclide))
if (micro_xs(p % event_nuclide) % absorption > ZERO .and. &
nuc % fissionable) then
score = p % absorb_wgt * &
associate (nuc => nuclides(p%event_nuclide))
if (micro_xs(p%event_nuclide)%absorption > ZERO .and. &
nuc%fissionable) then
score = p%absorb_wgt * &
nuc%reactions(nuc%index_fission(1))%Q_value * &
micro_xs(p % event_nuclide) % fission / &
micro_xs(p % event_nuclide) % absorption
micro_xs(p%event_nuclide)%fission / &
micro_xs(p%event_nuclide)%absorption
end if
end associate
else
@ -623,12 +623,12 @@ contains
! All fission events will contribute, so again we can use
! particle's weight entering the collision as the estimate for
! the fission energy production rate
associate (nuc => nuclides(p % event_nuclide))
if (nuc % fissionable) then
score = p % last_wgt * &
associate (nuc => nuclides(p%event_nuclide))
if (nuc%fissionable) then
score = p%last_wgt * &
nuc%reactions(nuc%index_fission(1))%Q_value * &
micro_xs(p % event_nuclide) % fission / &
micro_xs(p % event_nuclide) % absorption
micro_xs(p%event_nuclide)%fission / &
micro_xs(p%event_nuclide)%absorption
end if
end associate
end if
@ -636,7 +636,7 @@ contains
else
if (i_nuclide > 0) then
associate (nuc => nuclides(i_nuclide))
if (nuc % fissionable) then
if (nuc%fissionable) then
score = nuc%reactions(nuc%index_fission(1))%Q_value * &
micro_xs(i_nuclide)%fission * atom_density * flux
end if