diff --git a/src/ace.F90 b/src/ace.F90 index c6d9b02214..e97338b558 100644 --- a/src/ace.F90 +++ b/src/ace.F90 @@ -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 diff --git a/src/cross_section.F90 b/src/cross_section.F90 index 969a397f02..f874eb2a74 100644 --- a/src/cross_section.F90 +++ b/src/cross_section.F90 @@ -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 diff --git a/src/fission.F90 b/src/fission.F90 index 4c7613db32..a005d9f5bc 100644 --- a/src/fission.F90 +++ b/src/fission.F90 @@ -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 diff --git a/src/geometry.F90 b/src/geometry.F90 index 3f8b43d4cb..22f5c3a153 100644 --- a/src/geometry.F90 +++ b/src/geometry.F90 @@ -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 diff --git a/src/output.F90 b/src/output.F90 index 19ae6268be..c31b20b70b 100644 --- a/src/output.F90 +++ b/src/output.F90 @@ -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 diff --git a/src/tally.F90 b/src/tally.F90 index c4ca2028bd..09e30a2426 100644 --- a/src/tally.F90 +++ b/src/tally.F90 @@ -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