# 3. Settings Specification – settings.xml¶

All simulation parameters and miscellaneous options are specified in the settings.xml file.

## 3.1. <batches> Element¶

The <batches> element indicates the total number of batches to execute, where each batch corresponds to a tally realization. In a fixed source calculation, each batch consists of a number of source particles. In an eigenvalue calculation, each batch consists of one or many fission source iterations (generations), where each generation itself consists of a number of source neutrons.

Default: None

## 3.2. <confidence_intervals> Element¶

The <confidence_intervals> element has no attributes and has an accepted value of “true” or “false”. If set to “true”, uncertainties on tally results will be reported as the half-width of the 95% two-sided confidence interval. If set to “false”, uncertainties on tally results will be reported as the sample standard deviation.

Default: false

## 3.3. <create_fission_neutrons> Element¶

The <create_fission_neutrons> element indicates whether fission neutrons should be created or not. If this element is set to “true”, fission neutrons will be created; otherwise the fission is treated as capture and no fission neutron will be created. Note that this option is only applied to fixed source calculation. For eigenvalue calculation, fission will always be treated as real fission.

Default: true

## 3.4. <cutoff> Element¶

The <cutoff> element indicates two kinds of cutoffs. The first is the weight cutoff used below which particles undergo Russian roulette. Surviving particles are assigned a user-determined weight. Note that weight cutoffs and Russian rouletting are not turned on by default. The second is the energy cutoff which is used to kill particles under certain energy. The energy cutoff should not be used unless you know particles under the energy are of no importance to results you care. This element has the following attributes/sub-elements:

weight

The weight below which particles undergo Russian roulette.

Default: 0.25

weight_avg

The weight that is assigned to particles that are not killed after Russian roulette.

Default: 1.0

energy_neutron

The energy under which neutrons will be killed.

Default: 0.0

energy_photon

The energy under which photons will be killed.

Default: 1000.0

energy_electron

The energy under which electrons will be killed.

Default: 0.0

energy_positron

The energy under which positrons will be killed.

Default: 0.0

## 3.5. <delayed_photon_scaling>¶

Determines whether to scale the fission photon yield to account for delayed photon energy. The photon yields are scaled as (EGP + EGD)/EGP where EGP and EGD are the prompt and delayed photon components of energy release, respectively, from MF=1, MT=458 on an ENDF evaluation.

Default: true

## 3.6. <electron_treatment> Element¶

When photon transport is enabled, the <electron_treatment> element tells OpenMC whether to deposit all energy from electrons locally (led) or create secondary bremsstrahlung photons (ttb).

Default: ttb

## 3.7. <energy_mode> Element¶

The <energy_mode> element tells OpenMC if the run-mode should be continuous-energy or multi-group. Options for entry are: continuous-energy or multi-group.

Default: continuous-energy

## 3.8. <entropy_mesh> Element¶

The <entropy_mesh> element indicates the ID of a mesh that is to be used for calculating Shannon entropy. The mesh should cover all possible fissionable materials in the problem and is specified using a <mesh> Element.

## 3.9. <event_based>¶

Determines whether to use event-based parallelism instead of the default history-based parallelism.

Default: false

## 3.10. <generations_per_batch> Element¶

The <generations_per_batch> element indicates the number of total fission source iterations per batch for an eigenvalue calculation. This element is ignored for all run modes other than “eigenvalue”.

Default: 1

## 3.11. <inactive> Element¶

The <inactive> element indicates the number of inactive batches used in a k-eigenvalue calculation. In general, the starting fission source iterations in an eigenvalue calculation can not be used to contribute to tallies since the fission source distribution and eigenvalue are generally not converged immediately. This element is ignored for all run modes other than “eigenvalue”.

Default: 0

## 3.12. <keff_trigger> Element¶

The <keff_trigger> element (ignored for all run modes other than “eigenvalue”.) specifies a precision trigger on the combined $$k_{eff}$$. The trigger is a convergence criterion on the uncertainty of the estimated eigenvalue. It has the following attributes/sub-elements:

type

The type of precision trigger. Accepted options are “variance”, “std_dev”, and “rel_err”.

variance

Variance of the batch mean $$\sigma^2$$

std_dev

Standard deviation of the batch mean $$\sigma$$

rel_err

Relative error of the batch mean $$\frac{\sigma}{\mu}$$

Default: None

threshold

The precision trigger’s convergence criterion for the combined $$k_{eff}$$.

Default: None

Note

## 3.13. <log_grid_bins> Element¶

The <log_grid_bins> element indicates the number of bins to use for the logarithmic-mapped energy grid. Using more bins will result in energy grid searches over a smaller range at the expense of more memory. The default is based on the recommended value in LA-UR-14-24530.

Default: 8000

Note

This element is not used in the multi-group <energy_mode> Element.

## 3.14. <material_cell_offsets>¶

By default, OpenMC will count the number of instances of each cell filled with a material and generate “offset tables” that are used for cell instance tallies. The <material_cell_offsets> element allows a user to override this default setting and turn off the generation of offset tables, if desired, by setting it to false.

Default: true

## 3.15. <max_particles_in_flight> Element¶

This element indicates the number of neutrons to run in flight concurrently when using event-based parallelism. A higher value uses more memory, but may be more efficient computationally.

Default: 100000

## 3.16. <max_order> Element¶

The <max_order> element allows the user to set a maximum scattering order to apply to every nuclide/material in the problem. That is, if the data library has $$P_3$$ data available, but <max_order> was set to 1, then, OpenMC will only use up to the $$P_1$$ data.

Default: Use the maximum order in the data library

Note

This element is not used in the continuous-energy <energy_mode> Element.

## 3.17. <mesh> Element¶

The <mesh> element describes a mesh that is used either for calculating Shannon entropy, applying the uniform fission site method, or in tallies. For Shannon entropy meshes, the mesh should cover all possible fissionable materials in the problem. It has the following attributes/sub-elements:

id

A unique integer that is used to identify the mesh.

dimension

The number of mesh cells in the x, y, and z directions, respectively.

Default: If this tag is not present, the number of mesh cells is automatically determined by the code.

lower_left

The Cartesian coordinates of the lower-left corner of the mesh.

Default: None

upper_right

The Cartesian coordinates of the upper-right corner of the mesh.

Default: None

## 3.18. <no_reduce> Element¶

The <no_reduce> element has no attributes and has an accepted value of “true” or “false”. If set to “true”, all user-defined tallies and global tallies will not be reduced across processors in a parallel calculation. This means that the accumulate score in one batch on a single processor is considered as an independent realization for the tally random variable. For a problem with large tally data, this option can significantly improve the parallel efficiency.

Default: false

## 3.19. <output> Element¶

The <output> element determines what output files should be written to disk during the run. The sub-elements are described below, where “true” will write out the file and “false” will not.

summary

Writes out an HDF5 summary file describing all of the user input files that were read in.

Default: true

tallies

Write out an ASCII file of tally results.

Default: true

Note

The tally results will always be written to a binary/HDF5 state point file.

path

Absolute or relative path where all output files should be written to. The specified path must exist or else OpenMC will abort.

Default: Current working directory

## 3.20. <particles> Element¶

This element indicates the number of neutrons to simulate per fission source iteration when a k-eigenvalue calculation is performed or the number of particles per batch for a fixed source simulation.

Default: None

## 3.21. <photon_transport> Element¶

The <photon_transport> element determines whether photon transport is enabled. This element has no attributes or sub-elements and can be set to either “false” or “true”.

Default: false

## 3.22. <ptables> Element¶

The <ptables> element determines whether probability tables should be used in the unresolved resonance range if available. This element has no attributes or sub-elements and can be set to either “false” or “true”.

Default: true

Note

This element is not used in the multi-group <energy_mode> Element.

## 3.23. <resonance_scattering> Element¶

The resonance_scattering element indicates to OpenMC that a method be used to properly account for resonance elastic scattering (typically for nuclides with Z > 40). This element can contain one or more of the following attributes or sub-elements:

enable

Indicates whether a resonance elastic scattering method should be turned on. Accepts values of “true” or “false”.

Default: If the <resonance_scattering> element is present, “true”.

method

Which resonance elastic scattering method is to be applied: “rvs” (relative velocity sampling) or “dbrc” (Doppler broadening rejection correction). Descriptions of each of these methods are documented here.

Default: “rvs”

energy_min

The energy in eV above which the resonance elastic scattering method should be applied.

Default: 0.01 eV

energy_max

The energy in eV below which the resonance elastic scattering method should be applied.

Default: 1000.0 eV

nuclides

A list of nuclides to which the resonance elastic scattering method should be applied.

Default: If <resonance_scattering> is present but the <nuclides> sub-element is not given, the method is applied to all nuclides with 0 K elastic scattering data present.

Note

If the resonance_scattering element is not given, the free gas, constant cross section scattering model, which has historically been used by Monte Carlo codes to sample target velocities, is used to treat the target motion of all nuclides. If resonance_scattering is present, the constant cross section method is applied below energy_min and the target-at-rest (asymptotic) kernel is used above energy_max.

Note

This element is not used in the multi-group <energy_mode> Element.

## 3.24. <run_mode> Element¶

The <run_mode> element indicates which run mode should be used when OpenMC is executed. This element has no attributes or sub-elements and can be set to “eigenvalue”, “fixed source”, “plot”, “volume”, or “particle restart”.

Default: None

## 3.25. <seed> Element¶

The seed element is used to set the seed used for the linear congruential pseudo-random number generator.

Default: 1

## 3.26. <source> Element¶

The source element gives information on an external source distribution to be used either as the source for a fixed source calculation or the initial source guess for criticality calculations. Multiple <source> elements may be specified to define different source distributions. Each one takes the following attributes/sub-elements:

strength

The strength of the source. If multiple sources are present, the source strength indicates the relative probability of choosing one source over the other.

Default: 1.0

particle

The source particle type, either neutron or photon.

Default: neutron

file

If this attribute is given, it indicates that the source is to be read from a binary source file whose path is given by the value of this element. Note, the number of source sites needs to be the same as the number of particles simulated in a fission source generation.

Default: None

library

If this attribute is given, it indicates that the source is to be instantiated from an externally compiled source function. This source can be as complex as is required to define the source for your problem. The library has a few basic requirements:

• It must contain a class that inherits from openmc::Source;

• The class must implement a function called sample();

• There must be an openmc_create_source() function that creates the source as a unique pointer. This function can be used to pass parameters through to the source from the XML, if needed.

More documentation on how to build sources can be found in Custom Sources.

Default: None

parameters

If this attribute is given, it provides the parameters to pass through to the class generated using the library parameter . More documentation on how to build parametrized sources can be found in Custom Parameterized Sources.

Default: None

space

An element specifying the spatial distribution of source sites. This element has the following attributes:

type

The type of spatial distribution. Valid options are “box”, “fission”, “point”, “cartesian”, “cylindrical”, and “spherical”. A “box” spatial distribution has coordinates sampled uniformly in a parallelepiped. A “fission” spatial distribution samples locations from a “box” distribution but only locations in fissionable materials are accepted. A “point” spatial distribution has coordinates specified by a triplet. A “cartesian” spatial distribution specifies independent distributions of x-, y-, and z-coordinates. A “cylindrical” spatial distribution specifies independent distributions of r-, phi-, and z-coordinates where phi is the azimuthal angle and the origin for the cylindrical coordinate system is specified by origin. A “spherical” spatial distribution specifies independent distributions of r-, cos_theta-, and phi-coordinates where cos_theta is the cosine of the angle with respect to the z-axis, phi is the azimuthal angle, and the sphere is centered on the coordinate (x0,y0,z0).

Default: None

parameters

For a “box” or “fission” spatial distribution, parameters should be given as six real numbers, the first three of which specify the lower-left corner of a parallelepiped and the last three of which specify the upper-right corner. Source sites are sampled uniformly through that parallelepiped.

For a “point” spatial distribution, parameters should be given as three real numbers which specify the (x,y,z) location of an isotropic point source.

For an “cartesian” distribution, no parameters are specified. Instead, the x, y, and z elements must be specified.

For a “cylindrical” distribution, no parameters are specified. Instead, the r, phi, z, and origin elements must be specified.

For a “spherical” distribution, no parameters are specified. Instead, the r, theta, phi, and origin elements must be specified.

Default: None

x

For an “cartesian” distribution, this element specifies the distribution of x-coordinates. The necessary sub-elements/attributes are those of a univariate probability distribution (see the description in Univariate Probability Distributions).

y

For an “cartesian” distribution, this element specifies the distribution of y-coordinates. The necessary sub-elements/attributes are those of a univariate probability distribution (see the description in Univariate Probability Distributions).

z

For both “cartesian” and “cylindrical” distributions, this element specifies the distribution of z-coordinates. The necessary sub-elements/attributes are those of a univariate probability distribution (see the description in Univariate Probability Distributions).

r

For “cylindrical” and “spherical” distributions, this element specifies the distribution of r-coordinates (cylindrical radius and spherical radius, respectively). The necessary sub-elements/attributes are those of a univariate probability distribution (see the description in Univariate Probability Distributions).

theta

For a “spherical” distribution, this element specifies the distribution of theta-coordinates. The necessary sub-elements/attributes are those of a univariate probability distribution (see the description in Univariate Probability Distributions).

phi

For “cylindrical” and “spherical” distributions, this element specifies the distribution of phi-coordinates. The necessary sub-elements/attributes are those of a univariate probability distribution (see the description in Univariate Probability Distributions).

origin

For “cylindrical and “spherical” distributions, this element specifies the coordinates for the origin of the coordinate system.

angle

An element specifying the angular distribution of source sites. This element has the following attributes:

type

The type of angular distribution. Valid options are “isotropic”, “monodirectional”, and “mu-phi”. The angle of the particle emitted from a source site is isotropic if the “isotropic” option is given. The angle of the particle emitted from a source site is the direction specified in the reference_uvw element/attribute if “monodirectional” option is given. The “mu-phi” option produces directions with the cosine of the polar angle and the azimuthal angle explicitly specified.

Default: isotropic

reference_uvw

The direction from which the polar angle is measured. Represented by the x-, y-, and z-components of a unit vector. For a monodirectional distribution, this defines the direction of all sampled particles.

mu

An element specifying the distribution of the cosine of the polar angle. Only relevant when the type is “mu-phi”. The necessary sub-elements/attributes are those of a univariate probability distribution (see the description in Univariate Probability Distributions).

phi

An element specifying the distribution of the azimuthal angle. Only relevant when the type is “mu-phi”. The necessary sub-elements/attributes are those of a univariate probability distribution (see the description in Univariate Probability Distributions).

energy

An element specifying the energy distribution of source sites. The necessary sub-elements/attributes are those of a univariate probability distribution (see the description in Univariate Probability Distributions).

Default: Watt spectrum with $$a$$ = 0.988 MeV and $$b$$ = 2.249 MeV -1

write_initial

An element specifying whether to write out the initial source bank used at the beginning of the first batch. The output file is named “initial_source.h5”

Default: false

### 3.26.1. Univariate Probability Distributions¶

Various components of a source distribution involve probability distributions of a single random variable, e.g. the distribution of the energy, the distribution of the polar angle, and the distribution of x-coordinates. Each of these components supports the same syntax with an element whose tag signifies the variable and whose sub-elements/attributes are as follows:

type

The type of the distribution. Valid options are “uniform”, “discrete”, “tabular”, “maxwell”, “watt”, and “mixture”. The “uniform” option produces variates sampled from a uniform distribution over a finite interval. The “discrete” option produces random variates that can assume a finite number of values (i.e., a distribution characterized by a probability mass function). The “tabular” option produces random variates sampled from a tabulated distribution where the density function is either a histogram or linearly-interpolated between tabulated points. The “watt” option produces random variates is sampled from a Watt fission spectrum (only used for energies). The “maxwell” option produce variates sampled from a Maxwell fission spectrum (only used for energies). The “mixture” option produces samples from univariate sub-distributions with given probabilities.

Default: None

parameters

For a “uniform” distribution, parameters should be given as two real numbers $$a$$ and $$b$$ that define the interval $$[a,b]$$ over which random variates are sampled.

For a “powerlaw” distribution, parameters should be given as three real numbers $$a$$ and $$b$$ that define the interval $$[a,b]$$ over which random variates are sampled and $$n$$ that defines the exponent of the probability distribution $$p(x)=c x^n$$

For a “discrete” or “tabular” distribution, parameters provides the $$(x,p)$$ pairs defining the discrete/tabular distribution. All $$x$$ points are given first followed by corresponding $$p$$ points.

For a “watt” distribution, parameters should be given as two real numbers $$a$$ and $$b$$ that parameterize the distribution $$p(x) dx = c e^{-x/a} \sinh \sqrt{b \, x} dx$$.

For a “maxwell” distribution, parameters should be given as one real number $$a$$ that parameterizes the distribution $$p(x) dx = c x e^{-x/a} dx$$.

Note

The above format should be used even when using the multi-group <energy_mode> Element.

interpolation

For a “tabular” distribution, interpolation can be set to “histogram” or “linear-linear” thereby specifying how tabular points are to be interpolated.

Default: histogram

pair

For a “mixture” distribution, this element provides a distribution and its corresponding probability.

probability

An attribute or pair that provides the probability of a univariate distribution within a “mixture” distribution.

dist

This sub-element of a pair element provides information on the corresponding univariate distribution.

## 3.27. <state_point> Element¶

The <state_point> element indicates at what batches a state point file should be written. A state point file can be used to restart a run or to get tally results at any batch. The default behavior when using this tag is to write out the source bank in the state_point file. This behavior can be customized by using the <source_point> element. This element has the following attributes/sub-elements:

batches

A list of integers separated by spaces indicating at what batches a state point file should be written.

Default: Last batch only

## 3.28. <source_point> Element¶

The <source_point> element indicates at what batches the source bank should be written. The source bank can be either written out within a state point file or separately in a source point file. This element has the following attributes/sub-elements:

batches

A list of integers separated by spaces indicating at what batches a state point file should be written. It should be noted that if the separate attribute is not set to “true”, this list must be a subset of state point batches.

Default: Last batch only

separate

If this element is set to “true”, a separate binary source point file will be written. Otherwise, the source sites will be written in the state point directly.

Default: false

write

If this element is set to “false”, source sites are not written to the state point or source point file. This can substantially reduce the size of state points if large numbers of particles per batch are used.

Default: true

overwrite_latest

If this element is set to “true”, a source point file containing the source bank will be written out to a separate file named source.binary or source.h5 depending on if HDF5 is enabled. This file will be overwritten at every single batch so that the latest source bank will be available. It should be noted that a user can set both this element to “true” and specify batches to write a permanent source bank.

Default: false

## 3.29. <surf_source_read> Element¶

The <surf_source_read> element specifies a surface source file for OpenMC to read source bank for initializing histories. This element has the following attributes/sub-elements:

path

Absolute or relative path to a surface source file to read in source bank.

Default: surface_source.h5 in current working directory

## 3.30. <surf_source_write> Element¶

The <surf_source_write> element triggers OpenMC to bank particles crossing certain surfaces and write out the source bank in a separate file called surface_source.h5. This element has the following attributes/sub-elements:

surface_ids

A list of integers separated by spaces indicating the unique IDs of surfaces for which crossing particles will be banked.

Default: None

max_particles

An integer indicating the maximum number of particles to be banked on specified surfaces per processor. The size of source bank in surface_source.h5 is limited to this value times the number of processors.

Default: None

## 3.31. <survival_biasing> Element¶

The <survival_biasing> element has no attributes and has an accepted value of “true” or “false”. If set to “true”, this option will enable the use of survival biasing, otherwise known as implicit capture or absorption.

Default: false

## 3.32. <tabular_legendre> Element¶

The optional <tabular_legendre> element specifies how the multi-group Legendre scattering kernel is represented if encountered in a multi-group problem. Specifically, the options are to either convert the Legendre expansion to a tabular representation or leave it as a set of Legendre coefficients. Converting to a tabular representation will cost memory but can allow for a decrease in runtime compared to leaving as a set of Legendre coefficients. This element has the following attributes/sub-elements:

enable

This attribute/sub-element denotes whether or not the conversion of a Legendre scattering expansion to the tabular format should be performed or not. A value of “true” means the conversion should be performed, “false” means it will not.

Default: true

num_points

If the conversion is to take place the number of tabular points is required. This attribute/sub-element allows the user to set the desired number of points.

Default: 33

Note

This element is only used in the multi-group <energy_mode> Element.

## 3.33. <temperature_default> Element¶

The <temperature_default> element specifies a default temperature in Kelvin that is to be applied to cells in the absence of an explicit cell temperature or a material default temperature.

Default: 293.6 K

## 3.34. <temperature_method> Element¶

The <temperature_method> element has an accepted value of “nearest” or “interpolation”. A value of “nearest” indicates that for each cell, the nearest temperature at which cross sections are given is to be applied, within a given tolerance (see <temperature_tolerance> Element). A value of “interpolation” indicates that cross sections are to be linear-linear interpolated between temperatures at which nuclear data are present (see Temperature Treatment).

Default: “nearest”

## 3.35. <temperature_multipole> Element¶

The <temperature_multipole> element toggles the windowed multipole capability on or off. If this element is set to “True” and the relevant data is available, OpenMC will use the windowed multipole method to evaluate and Doppler broaden cross sections in the resolved resonance range. This override other methods like “nearest” and “interpolation” in the resolved resonance range.

Default: False

## 3.36. <temperature_range> Element¶

The <temperature_range> element specifies a minimum and maximum temperature in Kelvin above and below which cross sections should be loaded for all nuclides and thermal scattering tables. This can be used for multi-physics simulations where the temperatures might change from one iteration to the next.

Default: None

## 3.37. <temperature_tolerance> Element¶

The <temperature_tolerance> element specifies a tolerance in Kelvin that is to be applied when the “nearest” temperature method is used. For example, if a cell temperature is 340 K and the tolerance is 15 K, then the closest temperature in the range of 325 K to 355 K will be used to evaluate cross sections.

Default: 10 K

## 3.38. <trace> Element¶

The <trace> element can be used to print out detailed information about a single particle during a simulation. This element should be followed by three integers: the batch number, generation number, and particle number.

Default: None

## 3.39. <track> Element¶

The <track> element specifies particles for which OpenMC will output binary files describing particle position at every step of its transport. This element should be followed by triplets of integers. Each triplet describes one particle. The integers in each triplet specify the batch number, generation number, and particle number, respectively.

Default: None

## 3.40. <trigger> Element¶

OpenMC includes tally precision triggers which allow the user to define uncertainty thresholds on $$k_{eff}$$ in the <keff_trigger> subelement of settings.xml, and/or tallies in tallies.xml. When using triggers, OpenMC will run until it completes as many batches as defined by <batches>. At this point, the uncertainties on all tallied values are computed and compared with their corresponding trigger thresholds. If any triggers have not been met, OpenMC will continue until either all trigger thresholds have been satisfied or <max_batches> has been reached.

The <trigger> element provides an active “toggle switch” for tally precision trigger(s), the maximum number of batches and the batch interval. It has the following attributes/sub-elements:

active

This determines whether or not to use trigger(s). Trigger(s) are used when this tag is set to “true”.

max_batches

This describes the maximum number of batches allowed when using trigger(s).

Note

When max_batches is set, the number of batches shown in the <batches> element represents minimum number of batches to simulate when using the trigger(s).

batch_interval

This tag describes the number of batches in between convergence checks. OpenMC will check if the trigger has been reached at each batch defined by batch_interval after the minimum number of batches is reached.

Note

If this tag is not present, the batch_interval is predicted dynamically by OpenMC for each convergence check. The predictive model assumes no correlation between fission sources distributions from batch-to-batch. This assumption is reasonable for fixed source and small criticality calculations, but is very optimistic for highly coupled full-core reactor problems.

## 3.41. <ufs_mesh> Element¶

The <ufs_mesh> element indicates the ID of a mesh that is used for re-weighting source sites at every generation based on the uniform fission site methodology described in Kelly et al., “MC21 Analysis of the Nuclear Energy Agency Monte Carlo Performance Benchmark Problem,” Proceedings of Physor 2012, Knoxville, TN (2012). The mesh should cover all possible fissionable materials in the problem and is specified using a <mesh> Element.

## 3.42. <verbosity> Element¶

The <verbosity> element tells the code how much information to display to the standard output. A higher verbosity corresponds to more information being displayed. The text of this element should be an integer between between 1 and 10. The verbosity levels are defined as follows:

1

don’t display any output

2

only show OpenMC logo

3

all of the above + headers

4

all of the above + results

5

all of the above + file I/O

6

all of the above + timing statistics and initialization messages

7

all of the above + $$k$$ by generation

9

all of the above + indicate when each particle starts

10

all of the above + event information

Default: 7

## 3.43. <volume_calc> Element¶

The <volume_calc> element indicates that a stochastic volume calculation should be run at the beginning of the simulation. This element has the following sub-elements/attributes:

cells

The unique IDs of cells for which the volume should be estimated.

Default: None

samples

The number of samples used to estimate volumes.

Default: None

lower_left

The lower-left Cartesian coordinates of a bounding box that is used to sample points within.

Default: None

upper_right

The upper-right Cartesian coordinates of a bounding box that is used to sample points within.

Default: None

## 3.44. <weight_windows> Element¶

The <weight_windows> element specifies all necessary parameters for mesh-based weight windows. This element has the following sub-elements/attributes:

id

A unique integer that is used to identify the weight windows

mesh

ID of a mesh that is to be used for weight windows

Default: None

particle_type

The particle that the weight windows will apply to (e.g., ‘neutron’)

Default: None

energy_bins

Monotonically increasing list of bounding energies in [eV] to be used for weight windows

Default: None

lower_ww_bounds

Lower weight window bound for each (energy bin, mesh bin) combination.

Default: None

upper_ww_bounds

Upper weight window bound for each (energy bin, mesh bin) combination.

Default: None

survival

The ratio of survival weight and lower weight window bound.

Default: 3.0

max_lower_bound_ratio

Maximum allowed ratio of a particle’s weight to the weight window’s lower bound. A factor will be applied to raise the weight window to be lower than the particle’s weight by a factor of max_lower_bound_ratio during transport if exceeded.

max_split

Maximum allowable number of particles when splitting

Default: 10

weight_cutoff

Threshold below which particles will be terminated

Default: $$10^{-38}$$