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Don't pre-compute kappa-fission cross sections.

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This also fixes a bug in the kappa-fission score. Before, kappa-fission was
computed as Q*fission, but this was done before URR cross sections were
determined. Thus, if fission changed in calculate_urr_xs, this wasn't reflected
in the kappa-fission score. Now, since it is all done at tally-time, there is no
inconsistency.
This commit is contained in:
Paul Romano 2015-10-30 15:47:01 -05:00
parent a3199df9d9
commit 441fd4f00d
5 changed files with 59 additions and 66 deletions
Download patch file Download diff file Expand all files Collapse all files

2
src/ace_header.F90
View file

@ -258,7 +258,6 @@ module ace_header
real(8) :: absorption ! microscopic absorption xs
real(8) :: fission ! microscopic fission xs
real(8) :: nu_fission ! microscopic production xs
real(8) :: kappa_fission ! microscopic energy-released from fission
! Information for S(a,b) use
integer :: index_sab ! index in sab_tables (zero means no table)
@ -281,7 +280,6 @@ module ace_header
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
contains

13
src/cross_section.F90
View file

@ -41,7 +41,6 @@ contains
material_xs % absorption = ZERO
material_xs % fission = ZERO
material_xs % nu_fission = ZERO
material_xs % kappa_fission = ZERO
! Exit subroutine if material is void
if (p % material == MATERIAL_VOID) return
@ -125,10 +124,6 @@ contains
! Add contributions to material macroscopic nu-fission cross section
material_xs % nu_fission = material_xs % nu_fission + &
atom_density * micro_xs(i_nuclide) % nu_fission
! Add contributions to material macroscopic energy release from fission
material_xs % kappa_fission = material_xs % kappa_fission + &
atom_density * micro_xs(i_nuclide) % kappa_fission
end do
end subroutine calculate_xs
@ -216,7 +211,6 @@ contains
! Initialize nuclide cross-sections to zero
micro_xs(i_nuclide) % fission = ZERO
micro_xs(i_nuclide) % nu_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) &
@ -238,13 +232,6 @@ contains
! Calculate microscopic nuclide nu-fission cross section
micro_xs(i_nuclide) % nu_fission = (ONE - f) * nuc % nu_fission( &
i_grid) + f * nuc % nu_fission(i_grid+1)
! 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) % kappa_fission = &
nuc % reactions(nuc % index_fission(1)) % Q_value * &
micro_xs(i_nuclide) % fission
end if
! If there is S(a,b) data for this nuclide, we need to do a few

86
src/tally.F90
View file

@ -527,19 +527,11 @@ contains
d = t % filters(dg_filter) % int_bins(d_bin)
! Compute the yield for this delayed group
<<<<<<< HEAD
yield = yield_delayed(nuc, E, d)
yield = yield_delayed(nuclides(i_nuclide), E, d)
! Compute the score and tally to bin
score = micro_xs(i_nuclide) % fission * yield &
* nu_delayed(nuc, E) * atom_density * flux
=======
yield = yield_delayed(nuclides(i_nuclide), p % E, d)
! Compute the score and tally to bin
score = micro_xs(i_nuclide) % fission * yield &
* nu_delayed(nuclides(i_nuclide), p % E) * atom_density * flux
>>>>>>> Make Nuclide%reactions allocatable by using associate constructs
* nu_delayed(nuclides(i_nuclide), E) * atom_density * flux
call score_fission_delayed_dg(t, d_bin, score, score_index)
end do
cycle SCORE_LOOP
@ -547,13 +539,8 @@ contains
! If the delayed group filter is not present, compute the score
! by multiplying the delayed-nu-fission macro xs by the flux
<<<<<<< HEAD
score = micro_xs(i_nuclide) % fission * nu_delayed(nuc, E)&
* atom_density * flux
=======
score = micro_xs(i_nuclide) % fission * &
nu_delayed(nuclides(i_nuclide), p % E) * atom_density * flux
>>>>>>> Make Nuclide%reactions allocatable by using associate constructs
nu_delayed(nuclides(i_nuclide), E) * atom_density * flux
end if
! Tally is on total nuclides
@ -578,19 +565,11 @@ contains
d = t % filters(dg_filter) % int_bins(d_bin)
! Get the yield for the desired nuclide and delayed group
<<<<<<< HEAD
yield = yield_delayed(nuc, E, d)
yield = yield_delayed(nuclides(i_nuc), E, d)
! Compute the score and tally to bin
score = micro_xs(i_nuc) % fission * yield &
* nu_delayed(nuc, E) * atom_density_ * flux
=======
yield = yield_delayed(nuclides(i_nuc), p % E, d)
! Compute the score and tally to bin
score = micro_xs(i_nuc) % fission * yield &
* nu_delayed(nuclides(i_nuc), p % E) * atom_density_ * flux
>>>>>>> Make Nuclide%reactions allocatable by using associate constructs
* nu_delayed(nuclides(i_nuc), E) * atom_density_ * flux
call score_fission_delayed_dg(t, d_bin, score, score_index)
end do
end do
@ -618,38 +597,67 @@ contains
case (SCORE_KAPPA_FISSION)
! Determine kappa-fission cross section on the fly. The ENDF standard
! (ENDF-102) states that MT 18 stores the fission energy as the Q_value
! (fission(1))
score = ZERO
if (t % estimator == ESTIMATOR_ANALOG) then
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 scale by kappa-fission
if (micro_xs(p % event_nuclide) % absorption > ZERO) then
score = p % absorb_wgt * &
micro_xs(p % event_nuclide) % kappa_fission / &
micro_xs(p % event_nuclide) % absorption
else
score = ZERO
end if
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
end if
end associate
else
! Skip any non-absorption events
if (p % event == EVENT_SCATTER) cycle SCORE_LOOP
! 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
score = p % last_wgt * &
micro_xs(p % event_nuclide) % kappa_fission / &
micro_xs(p % event_nuclide) % absorption
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
end if
end associate
end if
else
if (i_nuclide > 0) then
score = micro_xs(i_nuclide) % kappa_fission * atom_density * flux
associate (nuc => nuclides(i_nuclide))
if (nuc % fissionable) then
score = nuc%reactions(nuc%index_fission(1))%Q_value * &
micro_xs(i_nuclide)%fission * atom_density * flux
end if
end associate
else
score = material_xs % kappa_fission * flux
do l = 1, materials(p%material)%n_nuclides
! Determine atom density and index of nuclide
atom_density_ = materials(p%material)%atom_density(l)
i_nuc = materials(p%material)%nuclide(l)
! If nuclide is fissionable, accumulate kappa fission
associate(nuc => nuclides(i_nuc))
if (nuc % fissionable) then
score = score + nuc%reactions(nuc%index_fission(1))%Q_value * &
micro_xs(i_nuc)%fission * atom_density_ * flux
end if
end associate
end do
end if
end if
case (SCORE_EVENTS)
! Simply count number of scoring events
score = ONE

4
tests/test_many_scores/results_true.dat
View file

@ -33,8 +33,8 @@ tally 1:
7.620560E-01
1.816851E+00
1.102658E+00
1.338067E+02
5.986137E+03
1.337996E+02
5.985519E+03
2.247257E+01
1.683779E+02
1.512960E-01

20
tests/test_score_kappafission/results_true.dat
View file

@ -1,17 +1,17 @@
k-combined:
9.903196E-01 4.279617E-02
tally 1:
2.266048E+02
1.049833E+04
2.266169E+02
1.049923E+04
0.000000E+00
0.000000E+00
0.000000E+00
0.000000E+00
1.366139E+02
3.859561E+03
1.366590E+02
3.861651E+03
tally 2:
2.402814E+02
1.174003E+04
2.403775E+02
1.175130E+04
0.000000E+00
0.000000E+00
0.000000E+00
@ -19,11 +19,11 @@ tally 2:
1.270420E+02
3.297537E+03
tally 3:
2.217075E+02
1.003168E+04
2.217588E+02
1.003581E+04
0.000000E+00
0.000000E+00
0.000000E+00
0.000000E+00
1.375693E+02
3.872389E+03
1.376303E+02
3.875598E+03