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Merge remote-tracking branch 'upstream/develop' into hotfix-openc-openmc-lats

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This commit is contained in:
wbinventor@gmail.com 2016-02-01 16:41:25 -05:00
commit 3df9f3937b
6 changed files with 43 additions and 25 deletions
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3
docs/source/conf.py
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@ -154,7 +154,8 @@ html_title = "OpenMC Documentation"
# so a file named "default.css" will overwrite the builtin "default.css".
html_static_path = ['_static']
html_context = {'css_files': ['_static/theme_overrides.css']}
def setup(app):
app.add_stylesheet('theme_overrides.css')
# If not '', a 'Last updated on:' timestamp is inserted at every page bottom,
# using the given strftime format.

2
src/particle_header.F90
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@ -86,7 +86,7 @@ module particle_header
logical :: write_track = .false.
! Secondary particles created
integer :: n_secondary = 0
integer(8) :: n_secondary = 0
type(Bank) :: secondary_bank(MAX_SECONDARY)
contains

53
src/physics.F90
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@ -88,9 +88,14 @@ contains
! change when sampling fission sites. The following block handles all
! absorption (including fission)
if (nuc % fissionable .and. run_mode == MODE_EIGENVALUE) then
if (nuc % fissionable) then
call sample_fission(i_nuclide, i_reaction)
call create_fission_sites(p, i_nuclide, i_reaction)
if (run_mode == MODE_EIGENVALUE) then
call create_fission_sites(p, i_nuclide, i_reaction, fission_bank, n_bank)
elseif (run_mode == MODE_FIXEDSOURCE) then
call create_fission_sites(p, i_nuclide, i_reaction, &
p % secondary_bank, p % n_secondary)
end if
end if
! If survival biasing is being used, the following subroutine adjusts the
@ -1071,10 +1076,12 @@ contains
! neutrons produced from fission and creates appropriate bank sites.
!===============================================================================
subroutine create_fission_sites(p, i_nuclide, i_reaction)
subroutine create_fission_sites(p, i_nuclide, i_reaction, bank_array, size_bank)
type(Particle), intent(inout) :: p
integer, intent(in) :: i_nuclide
integer, intent(in) :: i_reaction
type(Bank), intent(inout) :: bank_array(:)
integer(8), intent(inout) :: size_bank
integer :: nu_d(MAX_DELAYED_GROUPS) ! number of delayed neutrons born
integer :: i ! loop index
@ -1123,27 +1130,35 @@ contains
nu = int(nu_t) + 1
end if
! Check for fission bank size getting hit
if (n_bank + nu > size(fission_bank)) then
if (master) call warning("Maximum number of sites in fission bank &
&reached. This can result in irreproducible results using different &
&numbers of processes/threads.")
! Check for bank size getting hit. For fixed source calculations, this is a
! fatal error. For eigenvalue calculations, it just means that k-effective
! was too high for a single batch.
if (size_bank + nu > size(bank_array)) then
if (run_mode == MODE_FIXEDSOURCE) then
call fatal_error("Secondary particle bank size limit reached. If you &
&are running a subcritical multiplication problem, k-effective &
&may be too close to one.")
else
if (master) call warning("Maximum number of sites in fission bank &
&reached. This can result in irreproducible results using different &
&numbers of processes/threads.")
end if
end if
! Bank source neutrons
if (nu == 0 .or. n_bank == size(fission_bank)) return
if (nu == 0 .or. size_bank == size(bank_array)) return
! Initialize counter of delayed neutrons encountered for each delayed group
! to zero.
nu_d(:) = 0
p % fission = .true. ! Fission neutrons will be banked
do i = int(n_bank,4) + 1, int(min(n_bank + nu, int(size(fission_bank),8)),4)
do i = int(size_bank,4) + 1, int(min(size_bank + nu, int(size(bank_array),8)),4)
! Bank source neutrons by copying particle data
fission_bank(i) % xyz = p % coord(1) % xyz
bank_array(i) % xyz = p % coord(1) % xyz
! Set weight of fission bank site
fission_bank(i) % wgt = ONE/weight
bank_array(i) % wgt = ONE/weight
! Sample cosine of angle -- fission neutrons are always emitted
! isotropically. Sometimes in ACE data, fission reactions actually have
@ -1153,17 +1168,17 @@ contains
! Sample azimuthal angle uniformly in [0,2*pi)
phi = TWO*PI*prn()
fission_bank(i) % uvw(1) = mu
fission_bank(i) % uvw(2) = sqrt(ONE - mu*mu) * cos(phi)
fission_bank(i) % uvw(3) = sqrt(ONE - mu*mu) * sin(phi)
bank_array(i) % uvw(1) = mu
bank_array(i) % uvw(2) = sqrt(ONE - mu*mu) * cos(phi)
bank_array(i) % uvw(3) = sqrt(ONE - mu*mu) * sin(phi)
! Sample secondary energy distribution for fission reaction and set energy
! in fission bank
fission_bank(i) % E = sample_fission_energy(nuc, nuc%reactions(&
i_reaction), p)
bank_array(i) % E = sample_fission_energy(nuc, &
nuc % reactions(i_reaction), p)
! Set the delayed group of the neutron
fission_bank(i) % delayed_group = p % delayed_group
bank_array(i) % delayed_group = p % delayed_group
! Increment the number of neutrons born delayed
if (p % delayed_group > 0) then
@ -1172,7 +1187,7 @@ contains
end do
! increment number of bank sites
n_bank = min(n_bank + nu, int(size(fission_bank),8))
size_bank = min(size_bank + nu, int(size(bank_array),8))
! Store total and delayed weight banked for analog fission tallies
p % n_bank = nu

1
src/tracking.F90
View file

@ -202,6 +202,7 @@ contains
if (p % n_secondary > 0) then
call p % initialize_from_source(p % secondary_bank(p % n_secondary))
p % n_secondary = p % n_secondary - 1
n_event = 0
! Enter new particle in particle track file
if (p % write_track) call add_particle_track()

1
tests/test_fixed_source/materials.xml
View file

@ -4,6 +4,7 @@
<material id="1">
<density value="7.5" units="g/cc" />
<nuclide name="O-16" xs="71c" ao="1.0" />
<nuclide name="U-238" xs="71c" ao="0.0001" />
</material>
</materials>

8
tests/test_fixed_source/results_true.dat
View file

@ -1,6 +1,6 @@
tally 1:
4.538791E+02
2.073271E+04
4.563929E+02
2.091711E+04
leakage:
9.830000E+00
9.663900E+00
9.780000E+00
9.566400E+00