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Switched to using kappa-fission to refer to the kappa-fission tally-related quantities instead of just k-fission. Code and Manual are updated to reflect this.

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Adam Nelson 2013-01-18 09:29:55 -05:00
parent e0978a80ea
commit 930f478532
8 changed files with 66 additions and 66 deletions
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9
docs/source/usersguide/input.rst
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@ -837,14 +837,13 @@ The following responses can be tallied.
:nu-fission:
Total production of neutrons due to fission
:k-fission:
Kappa-Fission; the recoverable energy production due to fission. The recoverable
:kappa-fission:
The recoverable energy production rate due to fission. The recoverable
energy is defined as the fission product kinetic energy, prompt and delayed neutron
kinetic energies, prompt and delayed :math:`\gamma`-ray total energies,
and the total energy released by the delayed :math:`\beta` particles. The
neutrino energy does not contribute to this response.
Note that the prompt and delayed :math:`\gamma`-rays are assumed to deposit
their energy locally. ``kappa-fission`` is also a valid identifier for this response.
neutrino energy does not contribute to this response. The prompt and delayed
:math:`\gamma`-rays are assumed to deposit their energy locally.
:events:
Number of scoring events

14
src/ace_header.F90
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@ -192,7 +192,7 @@ module ace_header
real(8) :: absorption ! microscopic absorption xs
real(8) :: fission ! microscopic fission xs
real(8) :: nu_fission ! microscopic production xs
real(8) :: k_fission ! microscopic energy-released from fission
real(8) :: kappa_fission ! microscopic energy-released from fission
! Information for S(a,b) use
logical :: use_sab ! in S(a,b) energy range?
@ -208,12 +208,12 @@ module ace_header
!===============================================================================
type MaterialMacroXS
real(8) :: total ! macroscopic total xs
real(8) :: elastic ! macroscopic elastic scattering xs
real(8) :: absorption ! macroscopic absorption xs
real(8) :: fission ! macroscopic fission xs
real(8) :: nu_fission ! macroscopic production xs
real(8) :: k_fission ! macroscopic energy-released from fission
real(8) :: total ! macroscopic total xs
real(8) :: elastic ! macroscopic elastic scattering xs
real(8) :: absorption ! macroscopic absorption xs
real(8) :: fission ! macroscopic fission xs
real(8) :: nu_fission ! macroscopic production xs
real(8) :: kappa_fission ! macroscopic energy-released from fission
end type MaterialMacroXS
end module ace_header

30
src/constants.F90
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@ -254,21 +254,21 @@ module constants
! Tally score type
integer, parameter :: N_SCORE_TYPES = 15
integer, parameter :: &
SCORE_FLUX = -1, & ! flux
SCORE_TOTAL = -2, & ! total reaction rate
SCORE_SCATTER = -3, & ! scattering rate
SCORE_NU_SCATTER = -4, & ! scattering production rate
SCORE_SCATTER_N = -5, & ! arbitrary scattering moment
SCORE_SCATTER_PN = -6, & ! system for scoring 0th through nth moment
SCORE_TRANSPORT = -7, & ! transport reaction rate
SCORE_DIFFUSION = -8, & ! diffusion coefficient
SCORE_N_1N = -9, & ! (n,1n) rate
SCORE_ABSORPTION = -10, & ! absorption rate
SCORE_FISSION = -11, & ! fission rate
SCORE_NU_FISSION = -12, & ! neutron production rate
SCORE_K_FISSION = -13, & ! fission energy production rate
SCORE_CURRENT = -14, & ! partial current
SCORE_EVENTS = -15 ! number of events
SCORE_FLUX = -1, & ! flux
SCORE_TOTAL = -2, & ! total reaction rate
SCORE_SCATTER = -3, & ! scattering rate
SCORE_NU_SCATTER = -4, & ! scattering production rate
SCORE_SCATTER_N = -5, & ! arbitrary scattering moment
SCORE_SCATTER_PN = -6, & ! system for scoring 0th through nth moment
SCORE_TRANSPORT = -7, & ! transport reaction rate
SCORE_DIFFUSION = -8, & ! diffusion coefficient
SCORE_N_1N = -9, & ! (n,1n) rate
SCORE_ABSORPTION = -10, & ! absorption rate
SCORE_FISSION = -11, & ! fission rate
SCORE_NU_FISSION = -12, & ! neutron production rate
SCORE_KAPPA_FISSION = -13, & ! fission energy production rate
SCORE_CURRENT = -14, & ! partial current
SCORE_EVENTS = -15 ! number of events
! Maximum scattering order supported
integer, parameter :: SCATT_ORDER_MAX = 10

12
src/cross_section.F90
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@ -33,7 +33,7 @@ contains
material_xs % absorption = ZERO
material_xs % fission = ZERO
material_xs % nu_fission = ZERO
material_xs % k_fission = ZERO
material_xs % kappa_fission = ZERO
! Exit subroutine if material is void
if (p % material == MATERIAL_VOID) return
@ -92,8 +92,8 @@ contains
atom_density * micro_xs(i_nuclide) % nu_fission
! Add contributions to material macroscopic energy release from fission
material_xs % k_fission = material_xs % k_fission + &
atom_density * micro_xs(i_nuclide) % k_fission
material_xs % kappa_fission = material_xs % kappa_fission + &
atom_density * micro_xs(i_nuclide) % kappa_fission
end do
end subroutine calculate_xs
@ -155,7 +155,7 @@ contains
! Initialize nuclide cross-sections to zero
micro_xs(i_nuclide) % fission = ZERO
micro_xs(i_nuclide) % nu_fission = ZERO
micro_xs(i_nuclide) % k_fission = ZERO
micro_xs(i_nuclide) % kappa_fission = ZERO
! Calculate microscopic nuclide total cross section
micro_xs(i_nuclide) % total = (ONE - f) * nuc % total(i_grid) &
@ -178,10 +178,10 @@ contains
micro_xs(i_nuclide) % nu_fission = (ONE - f) * nuc % nu_fission( &
i_grid) + f * nuc % nu_fission(i_grid+1)
! Calculate microscopic nuclide k-fission cross section
! Calculate microscopic nuclide kappa-fission cross section
! The ENDF standard (ENDF-102) states that MT 18 stores
! the fission energy as the Q_value (fission(1))
micro_xs(i_nuclide) % k_fission = &
micro_xs(i_nuclide) % kappa_fission = &
nuc % reactions(nuc % index_fission(1)) % Q_value * &
micro_xs(i_nuclide) % fission
end if

4
src/input_xml.F90
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@ -1852,8 +1852,8 @@ contains
! Set tally estimator to analog
t % estimator = ESTIMATOR_ANALOG
end if
case ('k-fission', 'kappa-fission')
t % score_bins(j) = SCORE_K_FISSION
case ('kappa-fission')
t % score_bins(j) = SCORE_KAPPA_FISSION
case ('current')
t % score_bins(j) = SCORE_CURRENT
t % type = TALLY_SURFACE_CURRENT

32
src/output.F90
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@ -815,8 +815,8 @@ contains
string = trim(string) // ' fission'
case (SCORE_NU_FISSION)
string = trim(string) // ' nu-fission'
case (SCORE_K_FISSION)
string = trim(string) // ' k-fission'
case (SCORE_KAPPA_FISSION)
string = trim(string) // ' kappa-fission'
case (SCORE_CURRENT)
string = trim(string) // ' current'
case default
@ -1444,20 +1444,20 @@ contains
filter_name(FILTER_ENERGYOUT) = "Outgoing Energy"
! Initialize names for scores
score_names(abs(SCORE_FLUX)) = "Flux"
score_names(abs(SCORE_TOTAL)) = "Total Reaction Rate"
score_names(abs(SCORE_SCATTER)) = "Scattering Rate"
score_names(abs(SCORE_NU_SCATTER)) = "Scattering Production Rate"
score_names(abs(SCORE_SCATTER_N)) = ""
score_names(abs(SCORE_SCATTER_PN)) = ""
score_names(abs(SCORE_TRANSPORT)) = "Transport Rate"
score_names(abs(SCORE_DIFFUSION)) = "Diffusion Coefficient"
score_names(abs(SCORE_N_1N)) = "(n,1n) Rate"
score_names(abs(SCORE_ABSORPTION)) = "Absorption Rate"
score_names(abs(SCORE_FISSION)) = "Fission Rate"
score_names(abs(SCORE_NU_FISSION)) = "Nu-Fission Rate"
score_names(abs(SCORE_K_FISSION)) = "Kappa-Fission Rate"
score_names(abs(SCORE_EVENTS)) = "Events"
score_names(abs(SCORE_FLUX)) = "Flux"
score_names(abs(SCORE_TOTAL)) = "Total Reaction Rate"
score_names(abs(SCORE_SCATTER)) = "Scattering Rate"
score_names(abs(SCORE_NU_SCATTER)) = "Scattering Production Rate"
score_names(abs(SCORE_SCATTER_N)) = ""
score_names(abs(SCORE_SCATTER_PN)) = ""
score_names(abs(SCORE_TRANSPORT)) = "Transport Rate"
score_names(abs(SCORE_DIFFUSION)) = "Diffusion Coefficient"
score_names(abs(SCORE_N_1N)) = "(n,1n) Rate"
score_names(abs(SCORE_ABSORPTION)) = "Absorption Rate"
score_names(abs(SCORE_FISSION)) = "Fission Rate"
score_names(abs(SCORE_NU_FISSION)) = "Nu-Fission Rate"
score_names(abs(SCORE_KAPPA_FISSION)) = "Kappa-Fission Rate"
score_names(abs(SCORE_EVENTS)) = "Events"
! Create filename for tally output
if (run_mode == MODE_TALLIES) then

29
src/tally.F90
View file

@ -320,13 +320,14 @@ contains
end if
end if
case (SCORE_K_FISSION)
case (SCORE_KAPPA_FISSION)
if (survival_biasing) then
! No fission events occur if survival biasing is on -- need to
! calculate fraction of absorptions that would have resulted in
! fission and multiply by Q
score = p % absorb_wgt * micro_xs(p % event_nuclide) % k_fission / &
score = p % absorb_wgt * &
micro_xs(p % event_nuclide) % kappa_fission / &
micro_xs(p % event_nuclide) % absorption
else
@ -570,8 +571,8 @@ contains
score = micro_xs(i_nuclide) % nu_fission * &
atom_density * flux
case (SCORE_K_FISSION)
score = micro_xs(i_nuclide) % k_fission * &
case (SCORE_KAPPA_FISSION)
score = micro_xs(i_nuclide) % kappa_fission * &
atom_density * flux
case (SCORE_EVENTS)
@ -648,8 +649,8 @@ contains
! Nu-fission cross section is pre-calculated
score = material_xs % nu_fission * flux
case (SCORE_K_FISSION)
score = material_xs % k_fission * flux
case (SCORE_KAPPA_FISSION)
score = material_xs % kappa_fission * flux
case (SCORE_EVENTS)
! For number of events, just score unity
@ -804,8 +805,8 @@ contains
case (SCORE_NU_FISSION)
score = micro_xs(i_nuclide) % nu_fission * atom_density * flux
case (SCORE_K_FISSION)
score = micro_xs(i_nuclide) % k_fission * atom_density * flux
case (SCORE_KAPPA_FISSION)
score = micro_xs(i_nuclide) % kappa_fission * atom_density * flux
case (SCORE_EVENTS)
score = ONE
@ -890,8 +891,8 @@ contains
case (SCORE_NU_FISSION)
score = material_xs % nu_fission * flux
case (SCORE_K_FISSION)
score = material_xs % k_fission * flux
case (SCORE_KAPPA_FISSION)
score = material_xs % kappa_fission * flux
case (SCORE_EVENTS)
score = ONE
@ -1213,8 +1214,8 @@ contains
case (SCORE_NU_FISSION)
score = micro_xs(i_nuclide) % nu_fission * &
atom_density * flux
case (SCORE_K_FISSION)
score = micro_xs(i_nuclide) % k_fission * atom_density * flux
case (SCORE_KAPPA_FISSION)
score = micro_xs(i_nuclide) % kappa_fission * atom_density * flux
case (SCORE_EVENTS)
score = ONE
case default
@ -1238,8 +1239,8 @@ contains
score = material_xs % fission * flux
case (SCORE_NU_FISSION)
score = material_xs % nu_fission * flux
case (SCORE_K_FISSION)
score = material_xs % k_fission * flux
case (SCORE_KAPPA_FISSION)
score = material_xs % kappa_fission * flux
case (SCORE_EVENTS)
score = ONE
case default

2
src/utils/statepoint.py
View file

@ -22,7 +22,7 @@ score_types = {-1: 'flux',
-10: 'absorption',
-11: 'fission',
-12: 'nu-fission',
-13: 'k-fission',
-13: 'kappa-fission',
-14: 'current',
-15: 'events',
1: '(n,total)',