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revised kappa-fission to work as a tracklength estimator (analog can still be accessed if the user wishes), and tested it for micros/macros, etc.
This commit is contained in:
Adam Nelson
parent
529e4ca5fc
commit
a0cdfc5c51
4 changed files with 67 additions and 12 deletions
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@ -192,6 +192,7 @@ module ace_header
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real(8) :: absorption ! microscopic absorption xs
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real(8) :: fission ! microscopic fission xs
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real(8) :: nu_fission ! microscopic production xs
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real(8) :: k_fission ! microscopic energy-released from fission
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! Information for S(a,b) use
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logical :: use_sab ! in S(a,b) energy range?
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@ -212,6 +213,7 @@ module ace_header
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real(8) :: absorption ! macroscopic absorption xs
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real(8) :: fission ! macroscopic fission xs
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real(8) :: nu_fission ! macroscopic production xs
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real(8) :: k_fission ! macroscopic energy-released from fission
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end type MaterialMacroXS
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end module ace_header
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@ -33,6 +33,7 @@ contains
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material_xs % absorption = ZERO
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material_xs % fission = ZERO
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material_xs % nu_fission = ZERO
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material_xs % k_fission = ZERO
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! Exit subroutine if material is void
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if (p % material == MATERIAL_VOID) return
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@ -89,7 +90,14 @@ contains
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! Add contributions to material macroscopic nu-fission cross section
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material_xs % nu_fission = material_xs % nu_fission + &
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atom_density * micro_xs(i_nuclide) % nu_fission
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! Add contributions to material macroscopic energy release from fission
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material_xs % k_fission = material_xs % k_fission + &
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atom_density * micro_xs(i_nuclide) % k_fission
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end do
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! renormalize k_fission
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!material_xs % k_fission = material_xs % k_fission / &
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! material_xs % fission
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end subroutine calculate_xs
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@ -150,6 +158,7 @@ contains
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! Initialize nuclide cross-sections to zero
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micro_xs(i_nuclide) % fission = ZERO
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micro_xs(i_nuclide) % nu_fission = ZERO
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micro_xs(i_nuclide) % k_fission = ZERO
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! Calculate microscopic nuclide total cross section
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micro_xs(i_nuclide) % total = (ONE - f) * nuc % total(i_grid) &
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@ -171,6 +180,22 @@ contains
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! Calculate microscopic nuclide nu-fission cross section
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micro_xs(i_nuclide) % nu_fission = (ONE - f) * nuc % nu_fission( &
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i_grid) + f * nuc % nu_fission(i_grid+1)
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! Calculate microscopic nuclide k-fission cross section
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! The ENDF standard (ENDF-102) states that MT 18 stores
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! the fission energy as the Q_value, so this will be obtained from
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! the MT of index_fission(1); the standard is mute on
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! whether or not the other fission channels store their own
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! fission energy output; after inspection of the ACE files, all
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! nuclides observed with multiple fission channels have the same
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! Q-value for all of them; therefore, this code just uses the first
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! fission reaction instead of storing the actual fission type which
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! occured.
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micro_xs(i_nuclide) % k_fission = &
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nuc % reactions(nuc % index_fission(1)) % Q_value * &
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micro_xs(i_nuclide) % fission
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end if
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! If there is S(a,b) data for this nuclide, we need to do a few
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@ -1855,9 +1855,6 @@ contains
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t % score_bins(j) = SCORE_NU_FISSION
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case ('k-fission')
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t % score_bins(j) = SCORE_K_FISSION
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! Set tally estimator to analog
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t % estimator = ESTIMATOR_ANALOG
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case ('chi')
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t % score_bins(j) = SCORE_CHI
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@ -323,19 +323,37 @@ contains
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! All fission events will contribute, so again we can use
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! particle's weight entering the collision as the estimate for
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! the fission energy production rate
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! The ENDF standard (ENDF-102) states that MT 18 stores
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! the fission energy as the Q_value, so this will be obtained from
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! the MT of index_fission(1); the standard is mute on
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! whether or not the other fission channels store their own
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! fission energy output; after inspection of the ACE files, all
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! nuclides observed with multiple fission channels have the same
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! Q-value for all of them; therefore, this code just uses the first
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! fission reaction instead of storing the actual fission type which
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! occured.
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n = nuclides(p % event_nuclide) % index_fission(1)
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score = last_wgt * &
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nuclides(p % event_nuclide) % reactions(n) % Q_value
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case (SCORE_CHI)
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! Skip any non-fission events
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if (p % event /= EVENT_FISSION) cycle SCORE_LOOP
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if (t % find_filter(FILTER_ENERGYOUT) > 0) then
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! Normally, we only need to make contributions to one scoring
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! bin. However, in the case of fission, since multiple
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! fission neutrons were emitted with different energies,
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! multiple outgoing energy bins may have been scored to. The
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! following logic treats this special case and results to
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! multiple bins
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call score_fission_eout(t, score_index)
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cycle SCORE_LOOP
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else
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! If there is no outgoing energy filter, than we only need to
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! score to one bin. For the score to be 'analog', we need to
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! score the number of particles that were banked in the
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! fission bank. Since this was weighted by 1/keff, we
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! multiply by keff to get the proper score.
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score = keff * p % wgt_bank
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end if
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case (SCORE_EVENTS)
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! Simply count number of scoring events
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score = ONE
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@ -543,6 +561,9 @@ contains
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case (SCORE_NU_FISSION)
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score = micro_xs(i_nuclide) % nu_fission * &
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atom_density * flux
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case (SCORE_K_FISSION)
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score = micro_xs(i_nuclide) % k_fission * &
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atom_density * flux
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case (SCORE_EVENTS)
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score = ONE
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case default
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@ -565,6 +586,8 @@ contains
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score = material_xs % fission * flux
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case (SCORE_NU_FISSION)
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score = material_xs % nu_fission * flux
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case (SCORE_K_FISSION)
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score = material_xs % k_fission * flux
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case (SCORE_EVENTS)
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score = ONE
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case default
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@ -660,6 +683,8 @@ contains
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score = micro_xs(i_nuclide) % fission * atom_density * flux
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case (SCORE_NU_FISSION)
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score = micro_xs(i_nuclide) % nu_fission * atom_density * flux
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case (SCORE_K_FISSION)
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score = micro_xs(i_nuclide) % k_fission * atom_density * flux
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case (SCORE_EVENTS)
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score = ONE
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case default
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@ -701,6 +726,8 @@ contains
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score = material_xs % fission * flux
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case (SCORE_NU_FISSION)
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score = material_xs % nu_fission * flux
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case (SCORE_K_FISSION)
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score = material_xs % k_fission * flux
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case (SCORE_EVENTS)
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score = ONE
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case default
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@ -966,6 +993,8 @@ contains
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case (SCORE_NU_FISSION)
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score = micro_xs(i_nuclide) % nu_fission * &
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atom_density * flux
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case (SCORE_K_FISSION)
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score = micro_xs(i_nuclide) % k_fission * atom_density * flux
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case (SCORE_EVENTS)
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score = ONE
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case default
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@ -989,6 +1018,8 @@ contains
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score = material_xs % fission * flux
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case (SCORE_NU_FISSION)
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score = material_xs % nu_fission * flux
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case (SCORE_K_FISSION)
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score = material_xs % k_fission * flux
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case (SCORE_EVENTS)
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score = ONE
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case default
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