8. Specifying Tallies¶
In order to obtain estimates of physical quantities in your simulation, you need
to create one or more tallies using the openmc.Tally
class. As
explained in detail in the theory manual, tallies
provide estimates of a scoring function times the flux integrated over some
region of phase space, as in:
Thus, to specify a tally, we need to specify what regions of phase space should be included when deciding whether to score an event as well as what the scoring function (\(f\) in the above equation) should be used. The regions of phase space are called filters and the scoring functions are simply called scores.
8.1. Filters¶
To specify the regions of phase space, one must create a
openmc.Filter
. Since openmc.Filter
is an abstract class, you
actually need to instantiate one of its sub-classes (for a full listing, see
Constructing Tallies). For example, to indicate that events that occur in a
given cell should score to the tally, we would create a
openmc.CellFilter
:
cell_filter = openmc.CellFilter([fuel.id, moderator.id, reflector.id])
Another commonly used filter is openmc.EnergyFilter
, which specifies
multiple energy bins over which events should be scored. Thus, if we wanted to
tally events where the incident particle has an energy in the ranges [0 eV, 4
eV] and [4 eV, 1 MeV], we would do the following:
energy_filter = openmc.EnergyFilter([0.0, 4.0, 1.0e6])
Energies are specified in eV and need to be monotonically increasing.
Caution
An energy bin between zero and the lowest energy specified is not included by default as it is in MCNP.
Once you have created a filter, it should be assigned to a openmc.Tally
instance through the Tally.filters
attribute:
tally.filters.append(cell_filter)
tally.filters.append(energy_filter)
# This is equivalent
tally.filters = [cell_filter, energy_filter]
Note
You are actually not required to assign any filters to a tally. If you create a tally with no filters, all events will score to the tally. This can be useful if you want to know, for example, a reaction rate over your entire model.
8.2. Scores¶
To specify the scoring functions, a list of strings needs to be given to the
Tally.scores
attribute. You can score the flux (‘flux’), a reaction rate
(‘total’, ‘fission’, etc.), or even scattering moments (e.g., ‘scatter-P3’). For
example, to tally the elastic scattering rate and the fission neutron
production, you’d assign:
tally.scores = ['elastic', 'nu-fission']
With no further specification, you will get the total elastic scattering rate
and the total fission neutron production. If you want reaction rates for a
particular nuclide or set of nuclides, you can set the Tally.nuclides
attribute to a list of strings indicating which nuclides. The nuclide names
should follow the same naming convention as that used
for material specification. If we wanted the reaction rates only for U235 and
U238 (separately), we’d set:
tally.nuclides = ['U235', 'U238']
You can also list ‘all’ as a nuclide which will give you a separate reaction rate for every nuclide in the model.
The following tables show all valid scores:
Score | Description |
---|---|
flux | Total flux. |
flux-YN | Spherical harmonic expansion of the direction of motion \(\left(\Omega\right)\) of the total flux. This score will tally all of the harmonic moments of order 0 to N. N must be between 0 and 10. |
Score | Description |
---|---|
absorption | Total absorption rate. This accounts for all reactions which do not produce secondary neutrons as well as fission. |
elastic | Elastic scattering reaction rate. |
fission | Total fission reaction rate. |
scatter | Total scattering rate. Can also be identified with the “scatter-0” response type. |
scatter-N | Tally the Nth scattering moment, where N
is the Legendre expansion order of the change in
particle angle \(\left(\mu\right)\). N must be
between 0 and 10. As an example, tallying the 2nd scattering moment would be specified as
<scores>scatter-2</scores> . |
scatter-PN | Tally all of the scattering moments from order 0 to
N, where N is the Legendre expansion order of the
change in particle angle
\(\left(\mu\right)\). That is, “scatter-P1” is
equivalent to requesting tallies of “scatter-0” and
“scatter-1”. Like for “scatter-N”, N must be
between 0 and 10. As an example, tallying up to the
2nd scattering moment would be specified
as <scores> scatter-P2 </scores> . |
scatter-YN | “scatter-YN” is similar to “scatter-PN” except an additional expansion is performed for the incoming particle direction \(\left(\Omega\right)\) using the real spherical harmonics. This is useful for performing angular flux moment weighting of the scattering moments. Like “scatter-PN”, “scatter-YN” will tally all of the moments from order 0 to N; N again must be between 0 and 10. |
total | Total reaction rate. |
total-YN | The total reaction rate expanded via spherical harmonics about the direction of motion of the neutron, \(\Omega\). This score will tally all of the harmonic moments of order 0 to N. N must be between 0 and 10. |
(n,2nd) | (n,2nd) reaction rate. |
(n,2n) | (n,2n) reaction rate. |
(n,3n) | (n,3n) reaction rate. |
(n,na) | (n,n\(\alpha\)) reaction rate. |
(n,n3a) | (n,n3\(\alpha\)) reaction rate. |
(n,2na) | (n,2n\(\alpha\)) reaction rate. |
(n,3na) | (n,3n\(\alpha\)) reaction rate. |
(n,np) | (n,np) reaction rate. |
(n,n2a) | (n,n2\(\alpha\)) reaction rate. |
(n,2n2a) | (n,2n2\(\alpha\)) reaction rate. |
(n,nd) | (n,nd) reaction rate. |
(n,nt) | (n,nt) reaction rate. |
(n,nHe-3) | (n,n3He) reaction rate. |
(n,nd2a) | (n,nd2\(\alpha\)) reaction rate. |
(n,nt2a) | (n,nt2\(\alpha\)) reaction rate. |
(n,4n) | (n,4n) reaction rate. |
(n,2np) | (n,2np) reaction rate. |
(n,3np) | (n,3np) reaction rate. |
(n,n2p) | (n,n2p) reaction rate. |
(n,n*X*) | Level inelastic scattering reaction rate. The X indicates what which inelastic level, e.g., (n,n3) is third-level inelastic scattering. |
(n,nc) | Continuum level inelastic scattering reaction rate. |
(n,gamma) | Radiative capture reaction rate. |
(n,p) | (n,p) reaction rate. |
(n,d) | (n,d) reaction rate. |
(n,t) | (n,t) reaction rate. |
(n,3He) | (n,3He) reaction rate. |
(n,a) | (n,\(\alpha\)) reaction rate. |
(n,2a) | (n,2\(\alpha\)) reaction rate. |
(n,3a) | (n,3\(\alpha\)) reaction rate. |
(n,2p) | (n,2p) reaction rate. |
(n,pa) | (n,p\(\alpha\)) reaction rate. |
(n,t2a) | (n,t2\(\alpha\)) reaction rate. |
(n,d2a) | (n,d2\(\alpha\)) reaction rate. |
(n,pd) | (n,pd) reaction rate. |
(n,pt) | (n,pt) reaction rate. |
(n,da) | (n,d\(\alpha\)) reaction rate. |
Arbitrary integer | An arbitrary integer is interpreted to mean the reaction rate for a reaction with a given ENDF MT number. |
Score | Description |
---|---|
delayed-nu-fission | Total production of delayed neutrons due to fission. |
prompt-nu-fission | Total production of prompt neutrons due to fission. |
nu-fission | Total production of neutrons due to fission. |
nu-scatter, nu-scatter-N, nu-scatter-PN, nu-scatter-YN | These scores are similar in functionality to their
scatter* equivalents except the total
production of neutrons due to scattering is scored
vice simply the scattering rate. This accounts for
multiplicity from (n,2n), (n,3n), and (n,4n)
reactions. |
Score | Description |
---|---|
current | Used in combination with a mesh filter: Partial currents on the boundaries of each cell in a mesh. It may not be used in conjunction with any other score. Only energy and mesh filters may be used. Used in combination with a surface filter: Net currents on any surface previously defined in the geometry. It may be used along with any other filter, except mesh filters. Surfaces can alternatively be defined with cell from and cell filters thereby resulting in tallying partial currents. Units are particles per source particle. |
events | Number of scoring events. Units are events per source particle. |
inverse-velocity | The flux-weighted inverse velocity where the velocity is in units of centimeters per second. |
kappa-fission | The recoverable energy production rate due to fission. The recoverable energy is defined as the fission product kinetic energy, prompt and delayed neutron kinetic energies, prompt and delayed \(\gamma\)-ray total energies, and the total energy released by the delayed \(\beta\) particles. The neutrino energy does not contribute to this response. The prompt and delayed \(\gamma\)-rays are assumed to deposit their energy locally. Units are eV per source particle. |
fission-q-prompt | The prompt fission energy production rate. This energy comes in the form of fission fragment nuclei, prompt neutrons, and prompt \(\gamma\)-rays. This value depends on the incident energy and it requires that the nuclear data library contains the optional fission energy release data. Energy is assumed to be deposited locally. Units are eV per source particle. |
fission-q-recoverable | The recoverable fission energy production rate. This energy comes in the form of fission fragment nuclei, prompt and delayed neutrons, prompt and delayed \(\gamma\)-rays, and delayed \(\beta\)-rays. This tally differs from the kappa-fission tally in that it is dependent on incident neutron energy and it requires that the nuclear data library contains the optional fission energy release data. Energy is assumed to be deposited locally. Units are eV per source paticle. |
decay-rate | The delayed-nu-fission-weighted decay rate where the decay rate is in units of inverse seconds. |