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 generally called filters and the scoring functions are simply called scores.
The only cases when filters do not correspond directly with the regions of phase space are when expansion functions are applied in the integrand, such as for Legendre expansions of the scattering kernel.
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 subclasses (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’), or a reaction
rate (‘total’, ‘fission’, etc.). For example, to tally the elastic scattering
rate and the fission neutron production, you’d assign:
tally.scores = ['elastic', 'nufission']
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. 
Score 
Description 

absorption 
Total absorption rate. For incident neutrons, this accounts for all reactions that do not produce secondary neutrons as well as fission. For incident photons, this includes photoelectric and pair production. 
elastic 
Elastic scattering reaction rate. 
fission 
Total fission reaction rate. 
scatter 
Total scattering rate. 
total 
Total reaction rate. 
(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,n3He) 
(n,n^{3}He) 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 thirdlevel 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,^{3}He) 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. 
coherentscatter 
Coherent (Rayleigh) scattering reaction rate. 
incoherentscatter 
Incoherent (Compton) scattering reaction rate. 
photoelectric 
Photoelectric absorption reaction rate. 
pairproduction 
Pair production 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 

delayednufission 
Total production of delayed neutrons due to fission. 
promptnufission 
Total production of prompt neutrons due to fission. 
nufission 
Total production of neutrons due to fission. 
nuscatter 
This score is similar in functionality to the

H1production 
Total production of H1. 
H2production 
Total production of H2 (deuterium). 
H3production 
Total production of H3 (tritium). 
He3production 
Total production of He3. 
He4production 
Total production of He4 (alpha particles). 
Score 
Description 

current 
Used in combination with a meshsurface 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 meshsurface 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. 
inversevelocity 
The fluxweighted inverse velocity where the velocity is in units of centimeters per second. 
heating 
Total nuclear heating in units of eV per source particle. For neutrons, this corresponds to MT=301 produced by NJOY’s HEATR module while for photons, this is tallied from either direct photon energy deposition (analog estimator) or pregenerated photon heating number. See Heating and Energy Deposition 
heatinglocal 
Total nuclear heating in units of eV per source particle assuming energy from secondary photons is deposited locally. Note that this score should only be used for incident neutrons. See Heating and Energy Deposition. 
kappafission 
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. 
fissionqprompt 
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. 
fissionqrecoverable 
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 kappafission 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. 
decayrate 
The delayednufissionweighted decay rate where the decay rate is in units of inverse seconds. 
damageenergy 
Damage energy production in units of eV per source particle. This corresponds to MT=444 produced by NJOY’s HEATR module. 
8.3. Normalization of Tally Results¶
As described in Scores, all tally scores are normalized per
source particle simulated. However, for analysis of a given system, we usually
want tally scores in a more natural unit. For example, neutron flux is often
reported in units of particles/cm^{2}s. For a fixed source simulation,
it is usually straightforward to convert units if the source rate is known. For
example, if the system being modeled includes a source that is emitting 10^{4} neutrons per second, the tally results just need to be multipled by 10^{4}. This can either be done manually or using the
openmc.Source.strength
attribute.
For a \(k\)eigenvalue calculation, normalizing tally results is not as
simple because the source rate is not actually known. Instead, we typically know
the system power, \(P\), which represents how much energy is deposited per
unit time. Most of this energy originates from fission, but a small percentage
also results from other reactions (e.g., photons emitted from \((n,\gamma)\)
reactions). The most rigorous method to normalize tally results is to run a
coupled neutronphoton calculation and tally the heating
score over the
entire system. This score provides the heating rate in units of [eV/source],
which we’ll denote \(H\). Then, calculate the heating rate in J/source as
Dividing the power by the observed heating rate then gives us a normalization factor that can be applied to other tallies:
Multiplying by the normalization factor and dividing by volume, we can then get the flux in typical units:
There are several slight variations on this procedure:
Run a neutrononly calculation and estimate the total heating using the
heatinglocal
score (this requires that your nuclear data has local heating data available, such as in the official data library at https://openmc.org. See Heating and Energy Deposition for more information.)Run a neutrononly calculation and use the
kappafission
orfissionqrecoverable
scores along with an estimate of the extra heating due to neutron capture reactions.Calculate the overall fission rate and then used a fixed Q value to estimate the heating rate.
Note that the only difference between these and the above procedures is in how \(H'\) is estimated.