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Merge pull request #718 from samuelshaner/decay-rates

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Delayed neutron precursor decay rate tally
This commit is contained in:
Paul Romano 2016-09-16 00:03:03 +02:00 • committed by GitHub
commit 4f95b4a7f6
22 changed files with 820 additions and 230 deletions
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2
.travis.yml
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@ -42,7 +42,7 @@ install: true
before_script:
- if [[ ! -e $HOME/nndc_hdf5/cross_sections.xml ]]; then
wget https://anl.box.com/shared/static/68b2yhu8e6mx1f6hnbzz9mxsgg42d9ls.xz -O - | tar -C $HOME -xvJ;
wget https://anl.box.com/shared/static/fouwc8lh9he2wc97kzq65u4rp8zt9sgq.xz -O - | tar -C $HOME -xvJ;
fi
- export OPENMC_CROSS_SECTIONS=$HOME/nndc_hdf5/cross_sections.xml

552
docs/source/pythonapi/examples/mdgxs-part-i.ipynb
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File diff suppressed because one or more lines are too long

1
docs/source/pythonapi/index.rst
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@ -298,6 +298,7 @@ Multi-delayed-group Cross Sections
openmc.mgxs.ChiDelayed
openmc.mgxs.DelayedNuFissionXS
openmc.mgxs.Beta
openmc.mgxs.DecayRate
Multi-group Cross Section Libraries
-----------------------------------

5
docs/source/usersguide/input.rst
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@ -1890,6 +1890,11 @@ The ``<tally>`` element accepts the following sub-elements:
| |deposited locally. Units are MeV per source |
| |paticle. |
+----------------------+---------------------------------------------------+
|decay-rate |The delayed-nu-fission-weighted decay rate where |
| |the decay rate is in units of inverse seconds. |
| |This score type is not used in the |
| |multi-group :ref:`energy_mode`. |
+----------------------+---------------------------------------------------+
.. note::
The ``analog`` estimator is actually identical to the ``collision``

2
openmc/data/neutron.py
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@ -478,7 +478,7 @@ class IncidentNeutron(EqualityMixin):
if len(rxs) > 0:
data.summed_reactions[mt_sum] = rx = Reaction(mt_sum)
for T in data.temperatures:
rx.xs[T] = Sum([rx.xs[T] for rx in rxs])
rx.xs[T] = Sum([rx_i.xs[T] for rx_i in rxs])
# Read unresolved resonance probability tables
if 'urr' in group:

4
openmc/data/reaction.py
View file

@ -107,7 +107,9 @@ def _get_fission_products(ace):
for group in range(n_group):
delayed_neutron = Product('neutron')
delayed_neutron.emission_mode = 'delayed'
delayed_neutron.decay_rate = ace.xss[idx]
# Convert units of inverse shakes to inverse seconds
delayed_neutron.decay_rate = ace.xss[idx] * 1.e8
group_probability = Tabulated1D.from_ace(ace, idx + 1)
if np.all(group_probability.y == group_probability.y[0]):

159
openmc/mgxs/mdgxs.py
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@ -20,7 +20,8 @@ if sys.version_info[0] >= 3:
# Supported cross section types
MDGXS_TYPES = ['delayed-nu-fission',
'chi-delayed',
'beta']
'beta',
'decay-rate']
# Maximum number of delayed groups, from src/constants.F90
MAX_DELAYED_GROUPS = 8
@ -213,7 +214,7 @@ class MDGXS(MGXS):
Parameters
----------
mdgxs_type : {'delayed-nu-fission', 'chi-delayed', 'beta'}
mdgxs_type : {'delayed-nu-fission', 'chi-delayed', 'beta', 'decay-rate'}
The type of multi-delayed-group cross section object to return
domain : openmc.Material or openmc.Cell or openmc.Universe or
openmc.Mesh
@ -249,6 +250,8 @@ class MDGXS(MGXS):
delayed_groups)
elif mdgxs_type == 'beta':
mdgxs = Beta(domain, domain_type, energy_groups, delayed_groups)
elif mdgxs_type == 'decay-rate':
mdgxs = DecayRate(domain, domain_type, energy_groups, delayed_groups)
mdgxs.by_nuclide = by_nuclide
mdgxs.name = name
@ -1574,3 +1577,155 @@ class Beta(MDGXS):
super(Beta, self)._compute_xs()
return self._xs_tally
class DecayRate(MDGXS):
r"""The decay rate for delayed neutron precursors.
This class can be used for both OpenMC input generation and tally data
post-processing to compute spatially-homogenized and energy-integrated
multi-group and multi-delayed group cross sections for multi-group
neutronics calculations. At a minimum, one needs to set the
:attr:`DecayRate.energy_groups` and :attr:`DecayRate.domain` properties.
Tallies for the flux and appropriate reaction rates over the specified
domain are generated automatically via the :attr:`DecayRate.tallies`
property, which can then be appended to a :class:`openmc.Tallies` instance.
For post-processing, the :meth:`MGXS.load_from_statepoint` will pull in the
necessary data to compute multi-group cross sections from a
:class:`openmc.StatePoint` instance. The derived multi-group cross section
can then be obtained from the :attr:`DecayRate.xs_tally` property.
For a spatial domain :math:`V`, energy group :math:`[E_g,E_{g-1}]`, and
delayed group :math:`d`, the decay rate is calculated as:
.. math::
\langle \lambda_d \nu^d \sigma_f \phi \rangle &= \int_{r \in V} dr
\int_{4\pi} d\Omega' \int_0^\infty dE' \int_0^\infty dE \; \lambda_d \nu^d
\sigma_f (r, E') \psi(r, E', \Omega') \\
\langle \nu^d \sigma_f \phi \rangle &= \int_{r \in V} dr \int_{4\pi}
d\Omega' \int_0^\infty dE' \int_0^\infty dE \; \chi(E) \nu^d
\sigma_f (r, E') \psi(r, E', \Omega') \\
\lambda_d &= \frac{\langle \lambda_d \nu^d \sigma_f \phi \rangle}
{\langle \nu^d \sigma_f \phi \rangle}
Parameters
----------
domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.Mesh
The domain for spatial homogenization
domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
The domain type for spatial homogenization
groups : openmc.mgxs.EnergyGroups
The energy group structure for energy condensation
by_nuclide : bool
If true, computes cross sections for each nuclide in domain
name : str, optional
Name of the multi-group cross section. Used as a label to identify
tallies in OpenMC 'tallies.xml' file.
delayed_groups : list of int
Delayed groups to filter out the xs
Attributes
----------
name : str, optional
Name of the multi-group cross section
rxn_type : str
Reaction type (e.g., 'total', 'nu-fission', etc.)
by_nuclide : bool
If true, computes cross sections for each nuclide in domain
domain : Material or Cell or Universe or Mesh
Domain for spatial homogenization
domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
Domain type for spatial homogenization
energy_groups : openmc.mgxs.EnergyGroups
Energy group structure for energy condensation
delayed_groups : list of int
Delayed groups to filter out the xs
tally_trigger : openmc.Trigger
An (optional) tally precision trigger given to each tally used to
compute the cross section
scores : list of str
The scores in each tally used to compute the multi-group cross section
filters : list of openmc.Filter
The filters in each tally used to compute the multi-group cross section
tally_keys : list of str
The keys into the tallies dictionary for each tally used to compute
the multi-group cross section
estimator : {'tracklength', 'analog'}
The tally estimator used to compute the multi-group cross section
tallies : collections.OrderedDict
OpenMC tallies needed to compute the multi-group cross section. The keys
are strings listed in the :attr:`DecayRate.tally_keys` property and
values are instances of :class:`openmc.Tally`.
rxn_rate_tally : openmc.Tally
Derived tally for the reaction rate tally used in the numerator to
compute the multi-group cross section. This attribute is None
unless the multi-group cross section has been computed.
xs_tally : openmc.Tally
Derived tally for the multi-group cross section. This attribute
is None unless the multi-group cross section has been computed.
num_subdomains : int
The number of subdomains is unity for 'material', 'cell' and 'universe'
domain types. When the This is equal to the number of cell instances
for 'distribcell' domain types (it is equal to unity prior to loading
tally data from a statepoint file).
num_nuclides : int
The number of nuclides for which the multi-group cross section is
being tracked. This is unity if the by_nuclide attribute is False.
nuclides : Iterable of str or 'sum'
The optional user-specified nuclides for which to compute cross
sections (e.g., 'U-238', 'O-16'). If by_nuclide is True but nuclides
are not specified by the user, all nuclides in the spatial domain
are included. This attribute is 'sum' if by_nuclide is false.
sparse : bool
Whether or not the MGXS' tallies use SciPy's LIL sparse matrix format
for compressed data storage
loaded_sp : bool
Whether or not a statepoint file has been loaded with tally data
derived : bool
Whether or not the MGXS is merged from one or more other MGXS
hdf5_key : str
The key used to index multi-group cross sections in an HDF5 data store
"""
def __init__(self, domain=None, domain_type=None, energy_groups=None,
delayed_groups=None, by_nuclide=False, name=''):
super(DecayRate, self).__init__(domain, domain_type, energy_groups,
delayed_groups, by_nuclide, name)
self._rxn_type = 'decay-rate'
self._estimator = 'analog'
@property
def scores(self):
return ['delayed-nu-fission', 'decay-rate']
@property
def tally_keys(self):
return ['delayed-nu-fission', 'decay-rate']
@property
def filters(self):
# Create the non-domain specific Filters for the Tallies
group_edges = self.energy_groups.group_edges
energy_filter = openmc.Filter('energy', group_edges)
if self.delayed_groups != None:
delayed_filter = openmc.Filter('delayedgroup', self.delayed_groups)
return [[delayed_filter, energy_filter], [delayed_filter, energy_filter]]
else:
return [[energy_filter], [energy_filter]]
@property
def xs_tally(self):
if self._xs_tally is None:
delayed_nu_fission = self.tallies['delayed-nu-fission']
# Compute the decay rate
self._xs_tally = self.rxn_rate_tally / delayed_nu_fission
super(DecayRate, self)._compute_xs()
return self._xs_tally

5
src/constants.F90
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@ -296,7 +296,7 @@ module constants
EVENT_ABSORB = 2
! Tally score type
integer, parameter :: N_SCORE_TYPES = 23
integer, parameter :: N_SCORE_TYPES = 24
integer, parameter :: &
SCORE_FLUX = -1, & ! flux
SCORE_TOTAL = -2, & ! total reaction rate
@ -320,7 +320,8 @@ module constants
SCORE_PROMPT_NU_FISSION = -20, & ! prompt neutron production rate
SCORE_INVERSE_VELOCITY = -21, & ! flux-weighted inverse velocity
SCORE_FISS_Q_PROMPT = -22, & ! prompt fission Q-value
SCORE_FISS_Q_RECOV = -23 ! recoverable fission Q-value
SCORE_FISS_Q_RECOV = -23, & ! recoverable fission Q-value
SCORE_DECAY_RATE = -24 ! delayed neutron precursor decay rate
! Maximum scattering order supported
integer, parameter :: MAX_ANG_ORDER = 10

2
src/endf.F90
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@ -40,6 +40,8 @@ contains
string = "fission"
case (SCORE_NU_FISSION)
string = "nu-fission"
case (SCORE_DECAY_RATE)
string = "decay-rate"
case (SCORE_DELAYED_NU_FISSION)
string = "delayed-nu-fission"
case (SCORE_PROMPT_NU_FISSION)

38
src/input_xml.F90
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@ -3230,22 +3230,6 @@ contains
! Check if total material was specified
if (trim(sarray(j)) == 'total') then
! Check if a delayedgroup filter is present for this tally
do l = 1, size(t % filters)
select type(filt => t % filters(l) % obj)
type is (DelayedGroupFilter)
call warning("A delayedgroup filter was used on a total &
&nuclide tally. Cross section libraries are not &
&guaranteed to have the same delayed group structure &
&across all isotopes. In particular, ENDF/B-VII.1 does &
&not have a consistent delayed group structure across &
&all isotopes while the JEFF 3.1.1 library has the same &
&delayed group structure across all isotopes. Use with &
&caution!")
end select
end do
t % nuclide_bins(j) = -1
cycle
end if
@ -3290,20 +3274,6 @@ contains
allocate(t % nuclide_bins(1))
t % nuclide_bins(1) = -1
t % n_nuclide_bins = 1
! Check if a delayedgroup filter is present for this tally
do l = 1, size(t % filters)
select type(filt => t % filters(l) % obj)
type is (DelayedGroupFilter)
call warning("A delayedgroup filter was used on a total nuclide &
&tally. Cross section libraries are not guaranteed to have the&
& same delayed group structure across all isotopes. In &
&particular, ENDF/B-VII.1 does not have a consistent delayed &
&group structure across all isotopes while the JEFF 3.1.1 &
&library has the same delayed group structure across all &
&isotopes. Use with caution!")
end select
end do
end if
! =======================================================================
@ -3417,8 +3387,9 @@ contains
end if
! Check if delayed group filter is used with any score besides
! delayed-nu-fission
if (score_name /= 'delayed-nu-fission' .and. &
! delayed-nu-fission or decay-rate
if ((score_name /= 'delayed-nu-fission' .and. &
score_name /= 'decay-rate') .and. &
t % find_filter(FILTER_DELAYEDGROUP) > 0) then
call fatal_error("Cannot tally " // trim(score_name) // " with a &
&delayedgroup filter.")
@ -3572,6 +3543,9 @@ contains
! Set tally estimator to analog
t % estimator = ESTIMATOR_ANALOG
end if
case ('decay-rate')
t % score_bins(j) = SCORE_DECAY_RATE
t % estimator = ESTIMATOR_ANALOG
case ('delayed-nu-fission')
t % score_bins(j) = SCORE_DELAYED_NU_FISSION
if (t % find_filter(FILTER_ENERGYOUT) > 0) then

1
src/output.F90
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@ -735,6 +735,7 @@ contains
score_names(abs(SCORE_NU_SCATTER_N)) = "Scattering Prod. Rate Moment"
score_names(abs(SCORE_NU_SCATTER_PN)) = "Scattering Prod. Rate Moment"
score_names(abs(SCORE_NU_SCATTER_YN)) = "Scattering Prod. Rate Moment"
score_names(abs(SCORE_DECAY_RATE)) = "Decay Rate"
score_names(abs(SCORE_DELAYED_NU_FISSION)) = "Delayed-Nu-Fission Rate"
score_names(abs(SCORE_PROMPT_NU_FISSION)) = "Prompt-Nu-Fission Rate"
score_names(abs(SCORE_INVERSE_VELOCITY)) = "Flux-Weighted Inverse Velocity"

137
src/tally.F90
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@ -94,6 +94,8 @@ contains
integer :: score_bin ! scoring bin, e.g. SCORE_FLUX
integer :: score_index ! scoring bin index
integer :: d ! delayed neutron index
integer :: g ! delayed neutron index
integer :: k ! loop index for bank sites
integer :: d_bin ! delayed group bin index
integer :: dg_filter ! index of delayed group filter
real(8) :: yield ! delayed neutron yield
@ -671,6 +673,141 @@ contains
end if
case (SCORE_DECAY_RATE)
! Set the delayedgroup filter index
dg_filter = t % find_filter(FILTER_DELAYEDGROUP)
if (survival_biasing) then
! No fission events occur if survival biasing is on -- need to
! calculate fraction of absorptions that would have resulted in
! delayed-nu-fission
if (micro_xs(p % event_nuclide) % absorption > ZERO .and. &
nuclides(p % event_nuclide) % fissionable) then
! Check if the delayed group filter is present
if (dg_filter > 0) then
select type(filt => t % filters(dg_filter) % obj)
type is (DelayedGroupFilter)
! Loop over all delayed group bins and tally to them
! individually
do d_bin = 1, filt % n_bins
! Get the delayed group for this bin
d = filt % groups(d_bin)
! Compute the yield for this delayed group
yield = nuclides(p % event_nuclide) &
% nu(E, EMISSION_DELAYED, d)
associate (rxn => nuclides(p % event_nuclide) % &
reactions(nuclides(p % event_nuclide) % index_fission(1)))
! Compute the score
score = p % absorb_wgt * yield * &
micro_xs(p % event_nuclide) % fission &
/ micro_xs(p % event_nuclide) % absorption &
* rxn % products(1 + d) % decay_rate
end associate
! Tally to bin
call score_fission_delayed_dg(t, d_bin, score, score_index)
end do
cycle SCORE_LOOP
end select
else
! If the delayed group filter is not present, compute the score
! by accumulating the absorbed weight times the decay rate times
! the fraction of the delayed-nu-fission xs to the absorption xs
! for all delayed groups.
score = ZERO
associate (rxn => nuclides(p % event_nuclide) % &
reactions(nuclides(p % event_nuclide) % index_fission(1)))
! We need to be careful not to overshoot the number of delayed
! groups since this could cause the range of the rxn % products
! array to be exceeded. Hence, we use the size of this array
! and not the MAX_DELAYED_GROUPS constant for this loop.
do d = 1, size(rxn % products) - 2
score = score + rxn % products(1 + d) % decay_rate * &
p % absorb_wgt * micro_xs(p % event_nuclide) % fission *&
nuclides(p % event_nuclide) % nu(E, EMISSION_DELAYED, d)&
/ micro_xs(p % event_nuclide) % absorption
end do
end associate
end if
end if
else
! Skip any non-fission events
if (.not. p % fission) cycle SCORE_LOOP
! If there is no outgoing energy filter, than we only need to
! score to one bin. For the score to be 'analog', we need to
! score the number of particles that were banked in the fission
! bank. Since this was weighted by 1/keff, we multiply by keff
! to get the proper score. Loop over the neutrons produced from
! fission and check which ones are delayed. If a delayed neutron is
! encountered, add its contribution to the fission bank to the
! score.
score = ZERO
! loop over number of particles banked
do k = 1, p % n_bank
! get the delayed group
g = fission_bank(n_bank - p % n_bank + k) % delayed_group
! Case for tallying delayed emissions
if (g /= 0) then
! Accumulate the decay rate times delayed nu fission score
associate (rxn => nuclides(p % event_nuclide) % &
reactions(nuclides(p % event_nuclide) % index_fission(1)))
! determine score based on bank site weight and keff.
score = score + keff * fission_bank(n_bank - p % n_bank + k) &
% wgt * rxn % products(1 + g) % decay_rate
end associate
! if the delayed group filter is present, tally to corresponding
! delayed group bin if it exists
if (dg_filter > 0) then
! declare the delayed group filter type
select type(filt => t % filters(dg_filter) % obj)
type is (DelayedGroupFilter)
! loop over delayed group bins until the corresponding bin is
! found
do d_bin = 1, filt % n_bins
d = filt % groups(d_bin)
! check whether the delayed group of the particle is equal to
! the delayed group of this bin
if (d == g) then
call score_fission_delayed_dg(t, d_bin, score, score_index)
end if
end do
end select
! Reset the score to zero
score = ZERO
end if
end if
end do
! If the delayed group filter is present, cycle because the
! score_fission_delayed_dg(...) has already tallied the score
if (dg_filter > 0) then
cycle SCORE_LOOP
end if
end if
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

2
tests/test_mgxs_library_condense/inputs_true.dat
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@ -1 +1 @@
2b2a2f5778b03f87d20fa6da96cc19db0489f69d4d5f0cac02dd407fd8e53a082081bfef35c4310cb2e0651254766d499e89279e711fd6fd93f913c964855839
723ce48152faa9ec1576c26a825dfa45cdd858e721bd698d72d559ac3b9e5c37e5048305bd9af9c3137a781798ab2eee20e5df46455bfd9d6819a08fd2bff47d

21
tests/test_mgxs_library_condense/results_true.dat
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@ -61,6 +61,13 @@
3 10000 4 1 total 0.002752 0.000240
4 10000 5 1 total 0.001231 0.000105
5 10000 6 1 total 0.000512 0.000044
material delayedgroup group in nuclide mean std. dev.
0 10000 1 1 total 0.000000 0.000000
1 10000 2 1 total 0.032739 0.028454
2 10000 3 1 total 0.120780 0.170809
3 10000 4 1 total 0.302780 0.109110
4 10000 5 1 total 0.000000 0.000000
5 10000 6 1 total 0.000000 0.000000
material group in nuclide mean std. dev.
0 10001 1 total 0.311594 0.013793
material group in nuclide mean std. dev.
@ -123,6 +130,13 @@
2 10001 3 1 total 0.0 0.0
3 10001 4 1 total 0.0 0.0
4 10001 5 1 total 0.0 0.0
5 10001 6 1 total 0.0 0.0
material delayedgroup group in nuclide mean std. dev.
0 10001 1 1 total 0.0 0.0
1 10001 2 1 total 0.0 0.0
2 10001 3 1 total 0.0 0.0
3 10001 4 1 total 0.0 0.0
4 10001 5 1 total 0.0 0.0
5 10001 6 1 total 0.0 0.0
material group in nuclide mean std. dev.
0 10002 1 total 0.904999 0.043964
@ -186,4 +200,11 @@
2 10002 3 1 total 0.0 0.0
3 10002 4 1 total 0.0 0.0
4 10002 5 1 total 0.0 0.0
5 10002 6 1 total 0.0 0.0
material delayedgroup group in nuclide mean std. dev.
0 10002 1 1 total 0.0 0.0
1 10002 2 1 total 0.0 0.0
2 10002 3 1 total 0.0 0.0
3 10002 4 1 total 0.0 0.0
4 10002 5 1 total 0.0 0.0
5 10002 6 1 total 0.0 0.0

2
tests/test_mgxs_library_distribcell/inputs_true.dat
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@ -1 +1 @@
ad427594bd8a68ad35382bc34b5932e7c78480b6e327caf63782d563fd6e6e5fb4e7b0dfd9094394a5db092f789c473dbb08cb6b3d0c0296edcfc247dbe95d6f
7d95d5cf3f4d1f3277d1a1bbc638b47888d7eb1e6afeedea73140ae5caeebead7c5f1a2945f2a18b49ce75adbd9df17773390530fee129d981a3af50141fe116

7
tests/test_mgxs_library_distribcell/results_true.dat
View file

@ -61,3 +61,10 @@
3 (0, 1, 2, 3, 4, 5, 6, 7, 8, 9, 10, 11, 12, 13,... 4 1 total 0.002727 0.000135
4 (0, 1, 2, 3, 4, 5, 6, 7, 8, 9, 10, 11, 12, 13,... 5 1 total 0.001210 0.000058
5 (0, 1, 2, 3, 4, 5, 6, 7, 8, 9, 10, 11, 12, 13,... 6 1 total 0.000504 0.000024
avg(distribcell) delayedgroup group in nuclide mean std. dev.
0 (0, 1, 2, 3, 4, 5, 6, 7, 8, 9, 10, 11, 12, 13,... 1 1 total 0.000000 0.000000
1 (0, 1, 2, 3, 4, 5, 6, 7, 8, 9, 10, 11, 12, 13,... 2 1 total 0.032739 0.046300
2 (0, 1, 2, 3, 4, 5, 6, 7, 8, 9, 10, 11, 12, 13,... 3 1 total 0.120780 0.170809
3 (0, 1, 2, 3, 4, 5, 6, 7, 8, 9, 10, 11, 12, 13,... 4 1 total 0.000000 0.000000
4 (0, 1, 2, 3, 4, 5, 6, 7, 8, 9, 10, 11, 12, 13,... 5 1 total 0.000000 0.000000
5 (0, 1, 2, 3, 4, 5, 6, 7, 8, 9, 10, 11, 12, 13,... 6 1 total 2.853000 4.034751

2
tests/test_mgxs_library_hdf5/inputs_true.dat
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@ -1 +1 @@
2b2a2f5778b03f87d20fa6da96cc19db0489f69d4d5f0cac02dd407fd8e53a082081bfef35c4310cb2e0651254766d499e89279e711fd6fd93f913c964855839
723ce48152faa9ec1576c26a825dfa45cdd858e721bd698d72d559ac3b9e5c37e5048305bd9af9c3137a781798ab2eee20e5df46455bfd9d6819a08fd2bff47d

39
tests/test_mgxs_library_hdf5/results_true.dat
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@ -111,6 +111,19 @@ domain=10000 type=beta
[ 3.21434855e-04 1.09939816e-03]
[ 1.82980497e-04 4.50738567e-04]
[ 7.48899920e-05 1.88812772e-04]]
domain=10000 type=decay-rate
[[ 0. 0. ]
[ 0. 0.032739]
[ 0. 0.12078 ]
[ 0. 0.30278 ]
[ 0. 0. ]
[ 0. 0. ]]
[[ 0. 0. ]
[ 0. 0.02845437]
[ 0. 0.17080871]
[ 0. 0.10910951]
[ 0. 0. ]
[ 0. 0. ]]
domain=10001 type=total
[ 0.31373767 0.3008214 ]
[ 0.0155819 0.02805245]
@ -212,6 +225,19 @@ domain=10001 type=chi-delayed
[ 0. 0.]
[ 0. 0.]]
domain=10001 type=beta
[[ 0. 0.]
[ 0. 0.]
[ 0. 0.]
[ 0. 0.]
[ 0. 0.]
[ 0. 0.]]
[[ 0. 0.]
[ 0. 0.]
[ 0. 0.]
[ 0. 0.]
[ 0. 0.]
[ 0. 0.]]
domain=10001 type=decay-rate
[[ 0. 0.]
[ 0. 0.]
[ 0. 0.]
@ -337,3 +363,16 @@ domain=10002 type=beta
[ 0. 0.]
[ 0. 0.]
[ 0. 0.]]
domain=10002 type=decay-rate
[[ 0. 0.]
[ 0. 0.]
[ 0. 0.]
[ 0. 0.]
[ 0. 0.]
[ 0. 0.]]
[[ 0. 0.]
[ 0. 0.]
[ 0. 0.]
[ 0. 0.]
[ 0. 0.]
[ 0. 0.]]

2
tests/test_mgxs_library_mesh/inputs_true.dat
View file

@ -1 +1 @@
68c7695d7ae0367c59155eab05d3fe859cae895170f2cd32d4463af340a9a2035be16221b1eda8e2a1132bfe1b8163ca8d964fb1b23274c232fb10fd88274aea
88b1a1b52bcae466bab63038d03e9b2fb9f2a28aac3d73573ca00d5f0c116aa9e1c97341215b80266e8317a0549f11d9adcb8726c82fcc1741a5a730e10c8186

26
tests/test_mgxs_library_mesh/results_true.dat
View file

@ -208,3 +208,29 @@
21 2 2 1 4 1 total 0.002143 0.001028
22 2 2 1 5 1 total 0.001026 0.000492
23 2 2 1 6 1 total 0.000408 0.000196
mesh 1 delayedgroup group in nuclide mean std. dev.
x y z
0 1 1 1 1 1 total 0.00000 0.000000
1 1 1 1 2 1 total 0.00000 0.000000
2 1 1 1 3 1 total 0.00000 0.000000
3 1 1 1 4 1 total 0.30278 0.428196
4 1 1 1 5 1 total 0.00000 0.000000
5 1 1 1 6 1 total 0.00000 0.000000
6 1 2 1 1 1 total 0.00000 0.000000
7 1 2 1 2 1 total 0.00000 0.000000
8 1 2 1 3 1 total 0.00000 0.000000
9 1 2 1 4 1 total 0.00000 0.000000
10 1 2 1 5 1 total 0.00000 0.000000
11 1 2 1 6 1 total 0.00000 0.000000
12 2 1 1 1 1 total 0.00000 0.000000
13 2 1 1 2 1 total 0.00000 0.000000
14 2 1 1 3 1 total 0.00000 0.000000
15 2 1 1 4 1 total 0.00000 0.000000
16 2 1 1 5 1 total 0.00000 0.000000
17 2 1 1 6 1 total 0.00000 0.000000
18 2 2 1 1 1 total 0.00000 0.000000
19 2 2 1 2 1 total 0.00000 0.000000
20 2 2 1 3 1 total 0.00000 0.000000
21 2 2 1 4 1 total 0.00000 0.000000
22 2 2 1 5 1 total 0.00000 0.000000
23 2 2 1 6 1 total 0.00000 0.000000

2
tests/test_mgxs_library_no_nuclides/inputs_true.dat
View file

@ -1 +1 @@
2b2a2f5778b03f87d20fa6da96cc19db0489f69d4d5f0cac02dd407fd8e53a082081bfef35c4310cb2e0651254766d499e89279e711fd6fd93f913c964855839
723ce48152faa9ec1576c26a825dfa45cdd858e721bd698d72d559ac3b9e5c37e5048305bd9af9c3137a781798ab2eee20e5df46455bfd9d6819a08fd2bff47d

39
tests/test_mgxs_library_no_nuclides/results_true.dat
View file

@ -123,6 +123,19 @@
6 10000 4 2 total 0.011426 0.001099
8 10000 5 2 total 0.004684 0.000451
10 10000 6 2 total 0.001962 0.000189
material delayedgroup group in nuclide mean std. dev.
1 10000 1 1 total 0.000000 0.000000
3 10000 2 1 total 0.000000 0.000000
5 10000 3 1 total 0.000000 0.000000
7 10000 4 1 total 0.000000 0.000000
9 10000 5 1 total 0.000000 0.000000
11 10000 6 1 total 0.000000 0.000000
0 10000 1 2 total 0.000000 0.000000
2 10000 2 2 total 0.032739 0.028454
4 10000 3 2 total 0.120780 0.170809
6 10000 4 2 total 0.302780 0.109110
8 10000 5 2 total 0.000000 0.000000
10 10000 6 2 total 0.000000 0.000000
material group in nuclide mean std. dev.
1 10001 1 total 0.313738 0.015582
0 10001 2 total 0.300821 0.028052
@ -247,6 +260,19 @@
4 10001 3 2 total 0.0 0.0
6 10001 4 2 total 0.0 0.0
8 10001 5 2 total 0.0 0.0
10 10001 6 2 total 0.0 0.0
material delayedgroup group in nuclide mean std. dev.
1 10001 1 1 total 0.0 0.0
3 10001 2 1 total 0.0 0.0
5 10001 3 1 total 0.0 0.0
7 10001 4 1 total 0.0 0.0
9 10001 5 1 total 0.0 0.0
11 10001 6 1 total 0.0 0.0
0 10001 1 2 total 0.0 0.0
2 10001 2 2 total 0.0 0.0
4 10001 3 2 total 0.0 0.0
6 10001 4 2 total 0.0 0.0
8 10001 5 2 total 0.0 0.0
10 10001 6 2 total 0.0 0.0
material group in nuclide mean std. dev.
1 10002 1 total 0.664572 0.031215
@ -372,4 +398,17 @@
4 10002 3 2 total 0.0 0.0
6 10002 4 2 total 0.0 0.0
8 10002 5 2 total 0.0 0.0
10 10002 6 2 total 0.0 0.0
material delayedgroup group in nuclide mean std. dev.
1 10002 1 1 total 0.0 0.0
3 10002 2 1 total 0.0 0.0
5 10002 3 1 total 0.0 0.0
7 10002 4 1 total 0.0 0.0
9 10002 5 1 total 0.0 0.0
11 10002 6 1 total 0.0 0.0
0 10002 1 2 total 0.0 0.0
2 10002 2 2 total 0.0 0.0
4 10002 3 2 total 0.0 0.0
6 10002 4 2 total 0.0 0.0
8 10002 5 2 total 0.0 0.0
10 10002 6 2 total 0.0 0.0