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changed p % E to p % last_E for collision and analog tallies and changed neutron velocity to cm/s
This commit is contained in:
Sam Shaner
parent
f86910cc3b
commit
4d51fb6e5b
3 changed files with 30 additions and 14 deletions
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@ -1507,7 +1507,7 @@ The ``<tally>`` element accepts the following sub-elements:
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:inverse-velocity:
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The flux-weighted inverse velocity where the velocity is in units of
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meters per second.
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centimeters per second.
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.. note::
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The ``analog`` estimator is actually identical to the ``collision``
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@ -66,7 +66,7 @@ module constants
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MASS_NEUTRON_MEV = 939.565379_8, & ! mass of a neutron in MeV/c^2
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MASS_PROTON = 1.007276466812_8, & ! mass of a proton in amu
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AMU = 1.660538921e-27_8, & ! 1 amu in kg
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C_LIGHT = 2.99792458e8_8, & ! speed of light in m/s
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C_LIGHT = 2.99792458e10_8, & ! speed of light in cm/s
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N_AVOGADRO = 0.602214129_8, & ! Avogadro's number in 10^24/mol
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K_BOLTZMANN = 8.6173324e-11_8, & ! Boltzmann constant in MeV/K
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INFINITY = huge(0.0_8), & ! positive infinity
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@ -65,6 +65,7 @@ contains
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real(8) :: macro_total ! material macro total xs
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real(8) :: macro_scatt ! material macro scatt xs
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real(8) :: uvw(3) ! particle direction
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real(8) :: E ! particle energy
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type(Material), pointer :: mat
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type(Reaction), pointer :: rxn
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type(Nuclide), pointer :: nuc
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@ -128,6 +129,14 @@ contains
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case (SCORE_INVERSE_VELOCITY)
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! make sure the correct energy is used
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if (t % estimator == ESTIMATOR_TRACKLENGTH) then
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E = p % E
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else
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E = p % last_E
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end if
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if (t % estimator == ESTIMATOR_ANALOG) then
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! All events score to an inverse velocity bin. We actually use a
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! collision estimator in place of an analog one since there is no way
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@ -140,11 +149,11 @@ contains
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score = p % last_wgt
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end if
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score = score / material_xs % total &
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/ (sqrt(TWO * p % E / (MASS_NEUTRON_MEV)) * C_LIGHT)
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/ (sqrt(TWO * E / (MASS_NEUTRON_MEV)) * C_LIGHT)
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else
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! For inverse velocity, we need no cross section
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score = flux / (sqrt(TWO * p % E / (MASS_NEUTRON_MEV)) * C_LIGHT)
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! For inverse velocity, we don't need a cross section
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score = flux / (sqrt(TWO * E / (MASS_NEUTRON_MEV)) * C_LIGHT)
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end if
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@ -425,6 +434,13 @@ contains
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case (SCORE_DELAYED_NU_FISSION)
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! make sure the correct energy is used
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if (t % estimator == ESTIMATOR_TRACKLENGTH) then
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E = p % E
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else
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E = p % last_E
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end if
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! Set the delayedgroup filter index and the number of delayed group bins
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dg_filter = t % find_filter(FILTER_DELAYEDGROUP)
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@ -460,11 +476,11 @@ contains
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d = t % filters(dg_filter) % int_bins(d_bin)
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! Compute the yield for this delayed group
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yield = yield_delayed(nuc, p % E, d)
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yield = yield_delayed(nuc, E, d)
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! Compute the score and tally to bin
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score = p % absorb_wgt * yield * micro_xs(p % event_nuclide) &
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% fission * nu_delayed(nuc, p % E) / &
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% fission * nu_delayed(nuc, E) / &
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micro_xs(p % event_nuclide) % absorption
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call score_fission_delayed_dg(t, d_bin, score, score_index)
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end do
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@ -474,7 +490,7 @@ contains
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! by multiplying the absorbed weight by the fraction of the
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! delayed-nu-fission xs to the absorption xs
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score = p % absorb_wgt * micro_xs(p % event_nuclide) &
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% fission * nu_delayed(nuc, p % E) / &
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% fission * nu_delayed(nuc, E) / &
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micro_xs(p % event_nuclide) % absorption
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end if
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end if
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@ -528,11 +544,11 @@ contains
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d = t % filters(dg_filter) % int_bins(d_bin)
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! Compute the yield for this delayed group
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yield = yield_delayed(nuc, p % E, d)
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yield = yield_delayed(nuc, E, d)
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! Compute the score and tally to bin
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score = micro_xs(i_nuclide) % fission * yield &
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* nu_delayed(nuc, p % E) * atom_density * flux
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* nu_delayed(nuc, E) * atom_density * flux
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call score_fission_delayed_dg(t, d_bin, score, score_index)
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end do
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cycle SCORE_LOOP
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@ -540,7 +556,7 @@ contains
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! If the delayed group filter is not present, compute the score
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! by multiplying the delayed-nu-fission macro xs by the flux
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score = micro_xs(i_nuclide) % fission * nu_delayed(nuc, p % E)&
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score = micro_xs(i_nuclide) % fission * nu_delayed(nuc, E)&
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* atom_density * flux
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end if
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@ -572,11 +588,11 @@ contains
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nuc => nuclides(i_nuc)
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! Get the yield for the desired nuclide and delayed group
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yield = yield_delayed(nuc, p % E, d)
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yield = yield_delayed(nuc, E, d)
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! Compute the score and tally to bin
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score = micro_xs(i_nuc) % fission * yield &
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* nu_delayed(nuc, p % E) * atom_density_ * flux
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* nu_delayed(nuc, E) * atom_density_ * flux
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call score_fission_delayed_dg(t, d_bin, score, score_index)
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end do
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end do
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@ -596,7 +612,7 @@ contains
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! Accumulate the contribution from each nuclide
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score = score + micro_xs(i_nuc) % fission &
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* nu_delayed(nuclides(i_nuc), p % E) * atom_density_ * flux
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* nu_delayed(nuclides(i_nuc), E) * atom_density_ * flux
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end do
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end if
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end if
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