FileTransportMaterial

construction:Undocumented Class

The FileTransportMaterial has not been documented. The content listed below should be used as a starting point for documenting the class, which includes the typical automatic documentation associated with a MooseObject; however, what is contained is ultimately determined by what is necessary to make the documentation clear for users.

Provides the neutron group velocity ( $var element = document.getElementById("moose-equation-4d5caed2-7b44-4893-85b5-e89497ea70e5");katex.render("v_{g}", element, {displayMode:false,throwOnError:false});$ ), neutron group total cross-section ( $var element = document.getElementById("moose-equation-a65fa8fc-54a3-493f-bc32-7de022cbfaca");katex.render("\\Sigma_{t,g}", element, {displayMode:false,throwOnError:false});$ ), and the scattering cross-section moments ( $var element = document.getElementById("moose-equation-715de4ac-1b25-42fa-8449-e9c5f73e40e7");katex.render("\\Sigma_{s, g', g, l}", element, {displayMode:false,throwOnError:false});$ ) for transport problems. The material properties are imported from files provided by the user. The user many also provide volumetric neutron source moments through the input parameter system.

Overview

Example Input File Syntax

Input Parameters

file_nameThe file to extract cross-sections from. The path of the file should be relative to the input deck (.i) file.
C++ Type:std::string
Controllable:No
Description:The file to extract cross-sections from. The path of the file should be relative to the input deck (.i) file.
source_material_idThe material ID used by the cross-section source to identify cross-sections for different geometric regions.
C++ Type:std::string
Controllable:No
Description:The material ID used by the cross-section source to identify cross-sections for different geometric regions.

Required Parameters

add_heatingFalseWhether or not this material should provide fission heating cross-sections.
Default:False
C++ Type:bool
Controllable:No
Description:Whether or not this material should provide fission heating cross-sections.
blockThe list of blocks (ids or names) that this object will be applied
C++ Type:std::vector<SubdomainName>
Controllable:No
Description:The list of blocks (ids or names) that this object will be applied
boundaryThe list of boundaries (ids or names) from the mesh where this object applies
C++ Type:std::vector<BoundaryName>
Controllable:No
Description:The list of boundaries (ids or names) from the mesh where this object applies
computeTrueWhen false, MOOSE will not call compute methods on this material. The user must call computeProperties() after retrieving the MaterialBase via MaterialBasePropertyInterface::getMaterialBase(). Non-computed MaterialBases are not sorted for dependencies.
Default:True
C++ Type:bool
Controllable:No
Description:When false, MOOSE will not call compute methods on this material. The user must call computeProperties() after retrieving the MaterialBase via MaterialBasePropertyInterface::getMaterialBase(). Non-computed MaterialBases are not sorted for dependencies.
constant_onNONEWhen ELEMENT, MOOSE will only call computeQpProperties() for the 0th quadrature point, and then copy that value to the other qps.When SUBDOMAIN, MOOSE will only call computeQpProperties() for the 0th quadrature point, and then copy that value to the other qps. Evaluations on element qps will be skipped
Default:NONE
C++ Type:MooseEnum
Options:NONE, ELEMENT, SUBDOMAIN
Controllable:No
Description:When ELEMENT, MOOSE will only call computeQpProperties() for the 0th quadrature point, and then copy that value to the other qps.When SUBDOMAIN, MOOSE will only call computeQpProperties() for the 0th quadrature point, and then copy that value to the other qps. Evaluations on element qps will be skipped
cross_section_sourcedetectThe source which generated the cross-sections. Used for parsing different cross-section formats. If set to 'detect' the material will attempt to determine the source and parse accordingly.
Default:detect
C++ Type:MooseEnum
Options:detect, gnat, openmc
Controllable:No
Description:The source which generated the cross-sections. Used for parsing different cross-section formats. If set to 'detect' the material will attempt to determine the source and parse accordingly.
declare_suffixAn optional suffix parameter that can be appended to any declared properties. The suffix will be prepended with a '_' character.
C++ Type:MaterialPropertyName
Unit:(no unit assumed)
Controllable:No
Description:An optional suffix parameter that can be appended to any declared properties. The suffix will be prepended with a '_' character.
group_sourceThe external source moments for all energy groups.
C++ Type:std::vector<double>
Unit:(no unit assumed)
Controllable:No
Description:The external source moments for all energy groups.
has_fissionFalseWhether or not the transport system accounts for fission reactions.
Default:False
C++ Type:bool
Controllable:No
Description:Whether or not the transport system accounts for fission reactions.
is_diffusionFalseWhether or not the transport system is initializing a diffusion approximation scheme.
Default:False
C++ Type:bool
Controllable:No
Description:Whether or not the transport system is initializing a diffusion approximation scheme.
is_saafTrueWhether or not the transport system is initializing a SAAF scheme.
Default:True
C++ Type:bool
Controllable:No
Description:Whether or not the transport system is initializing a SAAF scheme.
num_groups0The number of energy groups the energy spectrum is divided into.
Default:0
C++ Type:unsigned int
Controllable:No
Description:The number of energy groups the energy spectrum is divided into.
particle_typeneutronThe type of particle to be consuming material property data.
Default:neutron
C++ Type:MooseEnum
Options:neutron, photon
Controllable:No
Description:The type of particle to be consuming material property data.
saaf_c1Stabilization parameter c for the SAAF-CFEM scheme. eta = 0 and c = 1 is equivalent to the SAAF approach with no void treatment. Use with caution.
Default:1
C++ Type:double
Unit:(no unit assumed)
Controllable:No
Description:Stabilization parameter c for the SAAF-CFEM scheme. eta = 0 and c = 1 is equivalent to the SAAF approach with no void treatment. Use with caution.
saaf_eta0.5Stabilization parameter eta for the SAAF-CFEM scheme. eta = 0 and c = 1 is equivalent to the SAAF approach with no void treatment. Use with caution.
Default:0.5
C++ Type:double
Unit:(no unit assumed)
Controllable:No
Description:Stabilization parameter eta for the SAAF-CFEM scheme. eta = 0 and c = 1 is equivalent to the SAAF approach with no void treatment. Use with caution.
source_anisotropy0The external source anisotropy of the medium.
Default:0
C++ Type:unsigned int
Controllable:No
Description:The external source anisotropy of the medium.
transport_systemName of the transport system which will consume the provided material properties. If one is not provided the first transport system will be used.
C++ Type:std::string
Controllable:No
Description:Name of the transport system which will consume the provided material properties. If one is not provided the first transport system will be used.

Optional Parameters

control_tagsAdds user-defined labels for accessing object parameters via control logic.
C++ Type:std::vector<std::string>
Controllable:No
Description:Adds user-defined labels for accessing object parameters via control logic.
enableTrueSet the enabled status of the MooseObject.
Default:True
C++ Type:bool
Controllable:Yes
Description:Set the enabled status of the MooseObject.
implicitTrueDetermines whether this object is calculated using an implicit or explicit form
Default:True
C++ Type:bool
Controllable:No
Description:Determines whether this object is calculated using an implicit or explicit form
seed0The seed for the master random number generator
Default:0
C++ Type:unsigned int
Controllable:No
Description:The seed for the master random number generator
use_displaced_meshFalseWhether or not this object should use the displaced mesh for computation. Note that in the case this is true but no displacements are provided in the Mesh block the undisplaced mesh will still be used.
Default:False
C++ Type:bool
Controllable:No
Description:Whether or not this object should use the displaced mesh for computation. Note that in the case this is true but no displacements are provided in the Mesh block the undisplaced mesh will still be used.

Advanced Parameters

output_propertiesList of material properties, from this material, to output (outputs must also be defined to an output type)
C++ Type:std::vector<std::string>
Controllable:No
Description:List of material properties, from this material, to output (outputs must also be defined to an output type)
outputsnone Vector of output names where you would like to restrict the output of variables(s) associated with this object
Default:none
C++ Type:std::vector<OutputName>
Controllable:No
Description:Vector of output names where you would like to restrict the output of variables(s) associated with this object

Outputs Parameters

prop_getter_suffixAn optional suffix parameter that can be appended to any attempt to retrieve/get material properties. The suffix will be prepended with a '_' character.
C++ Type:MaterialPropertyName
Unit:(no unit assumed)
Controllable:No
Description:An optional suffix parameter that can be appended to any attempt to retrieve/get material properties. The suffix will be prepended with a '_' character.
use_interpolated_stateFalseFor the old and older state use projected material properties interpolated at the quadrature points. To set up projection use the ProjectedStatefulMaterialStorageAction.
Default:False
C++ Type:bool
Controllable:No
Description:For the old and older state use projected material properties interpolated at the quadrature points. To set up projection use the ProjectedStatefulMaterialStorageAction.

Material Property Retrieval Parameters

Input Files

(examples/2D_multisystem_cgfem/test.i)
(examples/2D_containment_2_transient_fine_e/neutronics_combined.i)
(examples/2D_surface_current_cgfem/test.i)
(examples/2D_water_steady_cgfem/test.i)
(examples/2D_containment_transient_fine_e/neutronics.i)
(examples/2D_subcritical_reactor_1_group/neutronics_transport.i)
(examples/2D_subcritical_reactor_4_groups/neutronics.i)
(examples/2D_ray_tracing/subcritical_storage_uncollided.i)
(examples/2D_subcritical_1_group_cgfem/neutronics.i)
(examples/2D_subcritical_8_group_seg_cgfem/neutronics.i)
(examples/2D_containment_2_transient_fine_e/neutronics.i)
(examples/3D_subcritical_assembly/neutronics_transport.i)
(examples/2D_containment_2_transient_fine_e/uncollided.i)
(examples/2D_subcritical_reactor_1_group/neutronics_transport_eigen.i)
(examples/2D_diff_water_steady_cgfem/test.i)
(examples/2D_subcritical_1_group_cgfem/neutronics_uncollided.i)
(examples/2D_C5G7_full/neutronics.i)
(examples/2D_subcritical_reactor_4_groups/neutronics_uncollided.i)
(examples/2D_C5G7_simplified/neutronics.i)
(examples/2D_subcritical_reactor_4_groups/neutronics_eigen.i)

(examples/2D_multisystem_cgfem/test.i)

# A simple test case with a purely absorbing medium and a point source in the
# middle of the domain.

[Mesh]
  [domain]
    type = CartesianMeshGenerator
    dim = 2
    dx = '10.0'
    dy = '10.0'
    ix = '101'
    iy = '101'
    subdomain_id = '
      1'
  []
[]

[TransportSystems]
  [NeutronSN]
    scheme = saaf_cfem
    particle_type = neutron
    num_groups = 2
    max_anisotropy = 0
    angular_flux_names = neutron_angular_flux
    flux_moment_names = neutron_flux_moment

    order = FIRST
    family = LAGRANGE

    n_azimuthal = 3
    n_polar = 3

    vacuum_boundaries = 'left right top bottom'

    point_source_locations = '5.0 5.0 0.0'
    point_source_moments = '1.0 0.0'
    point_source_anisotropies = '0'
    scale_sources = true
  []

  [PhotonSN]
    scheme = saaf_cfem
    particle_type = Photon
    num_groups = 2
    max_anisotropy = 0
    angular_flux_names = photon_angular_flux
    flux_moment_names = photon_flux_moment

    order = FIRST
    family = LAGRANGE

    n_azimuthal = 3
    n_polar = 3

    vacuum_boundaries = 'left right top bottom'

    point_source_locations = '5.0 5.0 0.0'
    point_source_moments = '0.0 1.0'
    point_source_anisotropies = '0'
    scale_sources = true
  []
[]

[TransportMaterials]
  [Water1]
    type = FileTransportMaterial
    transport_system = NeutronSN
    file_name = 'macro_xs.xml'
    source_material_id = '1'
    block = '1'
  []

  [Water2]
    type = FileTransportMaterial
    transport_system = PhotonSN
    file_name = 'macro_xs.xml'
    source_material_id = '1'
    block = '1'
  []
[]

[Outputs]
  exodus = true
[]

[Executioner]
  type = Steady
  solve_type = PJFNK
  petsc_options_iname = '-pc_type -pc_hypre_type -ksp_gmres_restart'
  petsc_options_value = ' hypre    boomeramg      600'

  l_max_its = 50
  nl_rel_tol = 1e-12
[]

(examples/2D_containment_2_transient_fine_e/neutronics_combined.i)

[Mesh]
  [NeutronicsDomain]
    type = FileMeshGenerator
    file = Neutronics_Mesh.e
  []
  uniform_refine = 0
[]

[TransportSystems]
  [Neutron]
    num_groups = 8
    max_anisotropy = 0
    scheme = saaf_cfem
    particle_type = neutron

    flux_moment_names = 'Flux_Moment'
    scale_sources = true

    order = FIRST
    family = LAGRANGE

    is_conservative_transfer_src = true

    n_azimuthal = 5
    n_polar = 5

    vacuum_boundaries = 'vacuum'
    current_boundaries = 'barrel'

    uncollided_from_multi_app = 'Uncollided'
    use_conservative_uncollided_transfers = false
  []
[]

[TransportMaterials]
  [Stainless]
    type = FileTransportMaterial
    transport_system = Neutron
    file_name = 'data/xs_macro.xml'
    source_material_id = '6'
    block = stainless
  []
  [Air]
    type = FileTransportMaterial
    transport_system = Neutron
    file_name = 'data/xs_macro.xml'
    source_material_id = '5'
    block = air
  []
  [Concrete]
    type = FileTransportMaterial
    transport_system = Neutron
    file_name = 'data/xs_macro.xml'
    source_material_id = '7'
    block = concrete
  []
  [Water]
    type = FileTransportMaterial
    transport_system = Neutron
    file_name = 'data/water_macro_xs.xml'
    source_material_id = '1'
    block = light_water
  []
[]

[Executioner]
  type = Steady
  solve_type = PJFNK
  petsc_options_iname = '-pc_type -pc_hypre_type -ksp_gmres_restart'
  petsc_options_value = ' hypre    boomeramg      10'

  l_max_its = 100
  nl_abs_tol = 1e-12

  automatic_scaling = true
[]

[MultiApps]
  [Uncollided]
    type = FullSolveMultiApp
    input_files = 'uncollided.i'
    execute_on = INITIAL
  []
[]

[Outputs]
  exodus = true
  execute_on = 'TIMESTEP_END'
[]

(examples/2D_surface_current_cgfem/test.i)

[Mesh]
  [gen]
    type = GeneratedMeshGenerator
    dim = 2
    xmin = 0
    xmax = 10
    ymin = 0
    ymax = 10
    nx = 51
    ny = 51
  []
[]

[TransportSystems]
  [Neutron]
    scheme = saaf_cfem
    particle_type = neutron
    num_groups = 2
    max_anisotropy = 1

    flux_moment_names = 'Flux_Moment'

    order = FIRST
    family = LAGRANGE

    n_azimuthal = 5
    n_polar = 5

    vacuum_boundaries = 'bottom'
    current_boundaries = 'top left right'

    boundary_current_anisotropy = '1 1 1'
    boundary_currents = '1e3 0.0; 1e3 0.0; 1e3 0.0'
    scale_sources = true
  []
[]

[TransportMaterials]
  [Water]
    type = FileTransportMaterial
    transport_system = Neutron
    file_name = 'macro_xs.xml'
    source_material_id = '1'
  []
[]

[Executioner]
  type = Steady
  solve_type = PJFNK
  petsc_options_iname = '-pc_type -pc_hypre_type -ksp_gmres_restart'
  petsc_options_value = ' hypre    boomeramg      100'

  l_max_its = 50
  nl_rel_tol = 1e-12

  automatic_scaling = true
  off_diagonals_in_auto_scaling = true
  compute_scaling_once = false
[]

[Outputs]
  exodus = true
  execute_on = 'TIMESTEP_END'
[]

(examples/2D_water_steady_cgfem/test.i)

# A simple test case with a purely absorbing medium and a point source in the
# middle of the domain.

[Mesh]
  [domain]
    type = CartesianMeshGenerator
    dim = 2
    dx = '3.5 3.1 3.5'
    dy = '3.5 3.1 3.5'
    ix = '35 31 35'
    iy = '35 31 35'
    subdomain_id = '
      1 1 1
      1 2 1
      1 1 1'
  []
[]

[TransportSystems]
  [Neutron]
    num_groups = 2
    max_anisotropy = 1
    scheme = saaf_cfem
    particle_type = neutron

    order = FIRST
    family = LAGRANGE

    n_azimuthal = 5
    n_polar = 5

    vacuum_boundaries = 'left right top bottom'

    point_source_locations = '5.0 5.0 0.0'
    point_source_moments = '1e6 0.0'
    point_source_anisotropies = '0'
    scale_sources = true
  []
[]

[TransportMaterials]
  [Water]
    type = FileTransportMaterial
    transport_system = Neutron
    file_name = 'macro_xs.xml'
    source_material_id = '1'
  []
[]

[Executioner]
  type = Steady
  solve_type = PJFNK
  petsc_options_iname = '-pc_type -pc_hypre_type -ksp_gmres_restart'
  petsc_options_value = ' hypre    boomeramg      100'
  l_max_its = 50
  nl_rel_tol = 1e-12

  automatic_scaling = true
  off_diagonals_in_auto_scaling = true
  compute_scaling_once = false
[]

[Outputs]
  exodus = true
  execute_on = 'TIMESTEP_END'
[]

(examples/2D_containment_transient_fine_e/neutronics.i)

[Mesh]
  [NeutronicsDomain]
    type = FileMeshGenerator
    file = Full_Neutronics_3_Simple.e
  []
  uniform_refine = 0
[]

[TransportSystems]
  [Neutron]
    num_groups = 8
    max_anisotropy = 0
    scheme = saaf_cfem
    particle_type = neutron

    flux_moment_names = 'Flux_Moment'
    scale_sources = true

    order = FIRST
    family = LAGRANGE

    is_conservative_transfer_src = true

    n_azimuthal = 5
    n_polar = 5

    vacuum_boundaries = 'vacuum'
    current_boundaries = 'neutron'

    boundary_current_anisotropy = '1'
    boundary_currents = '
    7.534300e7 1.060700e7 3.250000e5 6.347148e9
    3.566577e9 0.0        0.0        0.0'
  []
[]

[TransportMaterials]
  [RPV]
    type = FileTransportMaterial
    transport_system = Neutron
    file_name = 'data/xs_macro.xml'
    source_material_id = '6'
    block = rpv
  []
  [Air]
    type = FileTransportMaterial
    transport_system = Neutron
    file_name = 'data/xs_macro.xml'
    source_material_id = '5'
    block = air
  []
  [Walls]
    type = FileTransportMaterial
    transport_system = Neutron
    file_name = 'data/xs_macro.xml'
    source_material_id = '7'
    block = walls
  []
[]

[Executioner]
  type = Steady
  solve_type = PJFNK
  petsc_options_iname = '-ksp_gmres_restart'
  petsc_options_value = ' 100'

  l_max_its = 100
  nl_abs_tol = 1e-12

  automatic_scaling = true
[]

[Outputs]
  exodus = true
[]

(examples/2D_subcritical_reactor_1_group/neutronics_transport.i)

[Mesh]
  [NeutronicsDomain]
    type = FileMeshGenerator
    file = Radiation_Mesh.e
  []
[]

[TransportSystems]
  [Neutron]
    num_groups = 1
    max_anisotropy = 1
    scheme = saaf_cfem
    particle_type = neutron

    flux_moment_names = 'Flux_Moment'
    debug_disable_fission = false

    order = FIRST
    family = LAGRANGE

    n_azimuthal = 5
    n_polar = 5

    vacuum_boundaries = 'vacuum'

    point_source_locations = '96.52 96.52 0.0'
    point_source_moments = '1e7'
    point_source_anisotropies = '0'
    scale_sources = true
  []
[]

[TransportMaterials]
  [Air]
    type = FileTransportMaterial
    transport_system = Neutron
    file_name = 'macro_xs.xml'
    source_material_id = '5'
    block = '2 7 8 9 ' #
  []

  [Cladding]
    type = FileTransportMaterial
    transport_system = Neutron
    file_name = 'macro_xs.xml'
    source_material_id = '7' #
    block = 4
  []

  [Fuel]
    type = FileTransportMaterial
    transport_system = Neutron
    file_name = 'macro_xs.xml'
    source_material_id = '6' #
    block = 3
  []

  [Graphite]
    type = FileTransportMaterial
    transport_system = Neutron
    file_name = 'macro_xs.xml'
    source_material_id = '8' #
    block = '10 11'
  []

  [Control]
    type = FileTransportMaterial
    transport_system = Neutron
    file_name = 'macro_xs.xml'
    source_material_id = '9' #
    block = 6
  []
[]

[Executioner]
  type = Steady
  solve_type = PJFNK
  petsc_options_iname = '-ksp_gmres_restart'
  petsc_options_value = ' 600'

  l_max_its = 50
  nl_abs_tol = 1e-12

  automatic_scaling = true
  off_diagonals_in_auto_scaling = true
  compute_scaling_once = false
[]

[Outputs]
  exodus = true
  execute_on = 'TIMESTEP_END'
[]

(examples/2D_subcritical_reactor_4_groups/neutronics.i)

# This input deck computes collided fluxes using the merged uncollided fluxes for a
# 2D version of the OTU subcritical assembly.

[Mesh]
  [NeutronicsDomain]
    type = FileMeshGenerator
    file = 2D_quarter_core.e
  []
  uniform_refine = 0
[]

[TransportSystems]
  [Neutron]
    num_groups = 4
    max_anisotropy = 0
    scheme = saaf_cfem
    particle_type = neutron
    flux_moment_names = 'Flux_Moment'
    debug_disable_fission = false

    order = FIRST
    family = LAGRANGE

    n_azimuthal = 3
    n_polar = 3

    vacuum_boundaries = 'vacuum'
    reflective_boundaries = 'reflective'

    uncollided_from_multi_app = 'Merge'
    use_conservative_uncollided_transfers = false
  []
[]

[TransportMaterials]
  [Air]
    type = FileTransportMaterial
    transport_system = Neutron
    file_name = 'macro_xs.xml'
    source_material_id = '5' #
    block = 'air_gap src_air'
  []

  [Cladding]
    type = FileTransportMaterial
    transport_system = Neutron
    file_name = 'macro_xs.xml'
    source_material_id = '7' #
    block = 'cladding'
  []

  [Fuel]
    type = FileTransportMaterial
    transport_system = Neutron
    file_name = 'macro_xs.xml'
    source_material_id = '6' #
    block = 'fuel'
  []

  [Graphite]
    type = FileTransportMaterial
    transport_system = Neutron
    file_name = 'macro_xs.xml'
    source_material_id = '8' #
    block = 'core_graphite reflector_graphite'
  []

  [Control]
    type = FileTransportMaterial
    transport_system = Neutron
    file_name = 'macro_xs.xml'
    source_material_id = '8' #
    block = 'control_or_graphite'
  []
[]

[Executioner]
  type = Steady
  solve_type = PJFNK
  petsc_options_iname = '-pc_type -pc_hypre_type -ksp_gmres_restart'
  petsc_options_value = ' hypre    boomeramg      600'

  l_max_its = 50
  nl_abs_tol = 1e-12

  automatic_scaling = true
  off_diagonals_in_auto_scaling = true
  compute_scaling_once = false
[]

[Outputs]
  exodus = true
  execute_on = 'TIMESTEP_END'
[]

[MultiApps]
  [Merge]
    type = FullSolveMultiApp
    app_type = GnatApp
    input_files = 'neutronics_merge.i'
    execute_on = INITIAL
  []
[]

(examples/2D_ray_tracing/subcritical_storage_uncollided.i)

[Mesh]
  [NeutronicsDomain]
    type = FileMeshGenerator
    file = Storage_Radiation_Mesh_Tris.e
  []
[]

[UncollidedFlux]
  [Neutron]
    uncollided_flux_treatment = ray-tracing
    num_groups = 1
    max_anisotropy = 1

    volumetric_source_blocks = fuel
    volumetric_source_moments = '40.5717'
    volumetric_source_anisotropies = '0'

    rt_n_polar = 4
  []
[]

[TransportMaterials]
  [Air]
    type = FileTransportMaterial
    transport_system = 'Neutron'
    file_name = 'macro_xs.xml'
    source_material_id = '5'
    block = air
  []
  [Cladding]
    type = FileTransportMaterial
    transport_system = 'Neutron'
    file_name = 'macro_xs.xml'
    source_material_id = '6'
    block = cladding
  []
  [Fuel]
    type = FileTransportMaterial
    transport_system = 'Neutron'
    file_name = 'macro_xs.xml'
    source_material_id = '7'
    block = fuel
  []
  [Wood]
    type = FileTransportMaterial
    transport_system = 'Neutron'
    file_name = 'macro_xs.xml'
    source_material_id = '8'
    block = wood
  []
[]

[Problem]
  type = FEProblem
  solve = false
  kernel_coverage_check = false
  material_coverage_check = false
[]

[Executioner]
  type = Steady
[]

[Outputs]
  exodus = true
  execute_on = 'TIMESTEP_END'
[]

(examples/2D_subcritical_1_group_cgfem/neutronics.i)

[Mesh]
  [NeutronicsDomain]
    type = FileMeshGenerator
    file = Radiation_Mesh.e
  []
[]

[TransportSystems]
  [Neutron]
    num_groups = 1
    max_anisotropy = 1
    scheme = saaf_cfem
    particle_type = neutron

    output_angular_fluxes = false
    flux_moment_names = 'Flux_Moment'

    order = FIRST
    family = LAGRANGE

    n_azimuthal = 4
    n_polar = 4

    vacuum_boundaries = 'vacuum'

    is_conservative_transfer_src = true

    uncollided_from_multi_app = 'Uncollided'
  []
[]

[TransportMaterials]
  [Air]
    type = FileTransportMaterial
    transport_system = Neutron
    file_name = 'data/macro_xs.xml'
    source_material_id = '5'
    block = air
  []
  [Cladding]
    type = FileTransportMaterial
    transport_system = Neutron
    file_name = 'data/macro_xs.xml'
    source_material_id = '6'
    block = cladding
  []
  [Fuel]
    type = FileTransportMaterial
    transport_system = Neutron
    file_name = 'data/macro_xs.xml'
    source_material_id = '7'
    block = fuel
  []
  [Wood]
    type = FileTransportMaterial
    transport_system = Neutron
    file_name = 'data/macro_xs.xml'
    source_material_id = '8'
    block = wood
  []
[]

[Executioner]
  type = Steady
  solve_type = PJFNK
  petsc_options_iname = '-pc_type -pc_hypre_type -ksp_gmres_restart'
  petsc_options_value = ' hypre    boomeramg      10'

  l_max_its = 50
  nl_abs_tol = 1e-12

  automatic_scaling = true
  off_diagonals_in_auto_scaling = true
  compute_scaling_once = false
[]

[MultiApps]
  [Uncollided]
    type = FullSolveMultiApp
    app_type = GnatApp
    input_files = 'neutronics_uncollided.i'
    execute_on = INITIAL
  []
[]

[Outputs]
  exodus = true
  execute_on = 'TIMESTEP_END'
[]

(examples/2D_subcritical_8_group_seg_cgfem/neutronics.i)

[Mesh]
  [NeutronicsDomain]
    type = FileMeshGenerator
    file = Radiation_Mesh.e
  []
[]

[TransportSystems]
  [Neutron]
    num_groups = 8
    max_anisotropy = 1
    scheme = saaf_cfem
    particle_type = neutron

    flux_moment_names = 'Flux_Moment'
    is_conservative_transfer_src = true

    order = FIRST
    family = LAGRANGE

    n_azimuthal = 4
    n_polar = 4

    vacuum_boundaries = 'vacuum'

    volumetric_source_blocks = fuel
    volumetric_source_moments = '
    40.5717 0.0 0.0 0.0
    0.0     0.0 0.0 0.0'
    volumetric_source_anisotropies = '0'
  []
[]

[TransportMaterials]
  [Air]
    type = FileTransportMaterial
    transport_system = Neutron
    file_name = 'data/macro_xs.xml'
    source_material_id = '5'
    block = air
  []
  [Cladding]
    type = FileTransportMaterial
    transport_system = Neutron
    file_name = 'data/macro_xs.xml'
    source_material_id = '6'
    block = cladding
  []
  [Fuel]
    type = FileTransportMaterial
    transport_system = Neutron
    file_name = 'data/macro_xs.xml'
    source_material_id = '7'
    block = fuel
  []
  [Wood]
    type = FileTransportMaterial
    transport_system = Neutron
    file_name = 'data/macro_xs.xml'
    source_material_id = '8'
    block = wood
  []
[]

[Executioner]
  type = Steady
  solve_type = PJFNK
  #petsc_options_iname = '-pc_type -pc_hypre_type -ksp_gmres_restart'
  #petsc_options_value = ' hypre    boomeramg      600'
  petsc_options_iname = '-ksp_gmres_restart'
  petsc_options_value = ' 600'

  l_max_its = 50
  nl_abs_tol = 1e-12

  automatic_scaling = true
  off_diagonals_in_auto_scaling = true
  compute_scaling_once = false
[]

[Outputs]
  exodus = true
  execute_on = 'TIMESTEP_END'
[]

(examples/2D_containment_2_transient_fine_e/neutronics.i)

[Mesh]
  [NeutronicsDomain]
    type = FileMeshGenerator
    file = Neutronics_Mesh.e
  []
  uniform_refine = 0
[]

[TransportSystems]
  [Neutron]
    num_groups = 8
    max_anisotropy = 0
    scheme = saaf_cfem
    particle_type = neutron

    flux_moment_names = 'Flux_Moment'
    scale_sources = true

    order = FIRST
    family = LAGRANGE

    is_conservative_transfer_src = true

    n_azimuthal = 5
    n_polar = 5

    vacuum_boundaries = 'vacuum'
    current_boundaries = 'barrel'

    boundary_current_anisotropy = '1'
    boundary_currents = '
    7.534300e7 1.060700e7 3.250000e5 6.347148e9
    3.566577e9 0.0        0.0        0.0'
  []
[]

[TransportMaterials]
  [Stainless]
    type = FileTransportMaterial
    transport_system = Neutron
    file_name = 'data/xs_macro.xml'
    source_material_id = '6'
    block = stainless
  []
  [Air]
    type = FileTransportMaterial
    transport_system = Neutron
    file_name = 'data/xs_macro.xml'
    source_material_id = '5'
    block = air
  []
  [Concrete]
    type = FileTransportMaterial
    transport_system = Neutron
    file_name = 'data/xs_macro.xml'
    source_material_id = '7'
    block = concrete
  []
  [Water]
    type = FileTransportMaterial
    transport_system = Neutron
    file_name = 'data/water_macro_xs.xml'
    source_material_id = '1'
    block = light_water
  []
[]

[Executioner]
  type = Steady
  solve_type = PJFNK
  petsc_options_iname = '-pc_type -pc_hypre_type -ksp_gmres_restart'
  petsc_options_value = ' hypre    boomeramg      10'

  l_max_its = 100
  nl_abs_tol = 1e-12

  automatic_scaling = true
[]

[Outputs]
  exodus = true
[]

(examples/3D_subcritical_assembly/neutronics_transport.i)

[GlobalParams]
  file_name = 'data/rodded/group_3_L_0.xml'
[]

[Mesh]
  [NeutronicsDomain]
    type = FileMeshGenerator
    file = mesh_out.e
  []
  uniform_refine = 0
[]

[TransportSystems]
  [Neutron]
    num_groups = 3
    max_anisotropy = 0
    scheme = saaf_cfem
    particle_type = neutron

    flux_moment_names = 'Flux_Moment'
    debug_disable_fission = false
    eigen = true
    constant_ic = '1.0'

    order = FIRST
    family = LAGRANGE

    n_azimuthal = 3
    n_polar = 3

    vacuum_boundaries = '10000'
    reflective_boundaries = '10001'

    #volumetric_source_blocks = '5'
    #volumetric_source_anisotropies = '0'
    #volumetric_source_moments = '1.0 0.0 0.0 0.0'
  []
[]

[TransportMaterials]
  [Air]
    type = FileTransportMaterial
    transport_system = Neutron
    source_material_id = '0'
    block = '2 6 106'
  []

  [Cladding]
    type = FileTransportMaterial
    transport_system = Neutron
    source_material_id = '1'
    block = '1'
  []

  [Control]
    type = FileTransportMaterial
    transport_system = Neutron
    source_material_id = '2'
    block = '5 105'
  []

  [Fuel]
    type = FileTransportMaterial
    transport_system = Neutron
    source_material_id = '3'
    block = '0 100'
    add_heating = true
  []

  [CoreGraphite]
    type = FileTransportMaterial
    transport_system = Neutron
    source_material_id = '4'
    block = '3 103'
  []

  [ReflectorGraphite]
    type = FileTransportMaterial
    transport_system = Neutron
    source_material_id = '5'
    block = '4 104'
  []
[]

[Postprocessors]
  [TotalFlux]
    type = TotalFluxPostProcessor
    num_groups = 3
    group_scalar_fluxes = 'Flux_Moment_1_0_0 Flux_Moment_2_0_0 Flux_Moment_3_0_0'
    execute_on = LINEAR
    outputs = 'csv'
  []
[]

[VectorPostprocessors]
  [EigenValues]
    type = Eigenvalues
    inverse_eigenvalue = true
  []
[]

[Executioner]
  type = Eigenvalue

  initial_eigenvalue = 1.2
  normal_factor = 1.0
  normalization = 'TotalFlux'
  free_power_iterations = 6

  solve_type = PJFNK
  #petsc_options_iname = '-pc_type -pc_hypre_type -ksp_gmres_restart'
  #petsc_options_value = ' hypre    boomeramg      600'
  petsc_options_iname = '-ksp_gmres_restart'
  petsc_options_value = ' 600'

  nl_abs_tol = 1e-12

  automatic_scaling = true
  off_diagonals_in_auto_scaling = true
  compute_scaling_once = false
[]

[Outputs]
  csv = true
  exodus = true
  execute_on = 'TIMESTEP_END'
[]

(examples/2D_containment_2_transient_fine_e/uncollided.i)

[Mesh]
  [NeutronicsDomain]
    type = FileMeshGenerator
    file = Neutronics_Mesh.e
  []
  uniform_refine = 2
[]

[UncollidedFlux]
  [Neutron]
    uncollided_flux_treatment = sasf
    num_groups = 8
    max_anisotropy = 0

    point_source_locations = '0.0 0.0 0.0'
    point_source_moments = '7.534300e0 1.060700e0 3.250000e-2 6.347148e2 3.566577e2 0.0 0.0 0.0'
    point_source_anisotropies = '0'

    sasf_near_source_boundary = 'barrel'
    sasf_vacuum_boundaries = 'vacuum'
    sasf_near_source_cross_sections = '0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0'

    is_conservative_transfer_src = false
  []
[]

[TransportMaterials]
  [Stainless]
    type = FileTransportMaterial
    transport_system = Neutron
    file_name = 'data/xs_macro.xml'
    source_material_id = '6'
    block = stainless
  []
  [Air]
    type = FileTransportMaterial
    transport_system = Neutron
    file_name = 'data/xs_macro.xml'
    source_material_id = '5'
    block = air
  []
  [Concrete]
    type = FileTransportMaterial
    transport_system = Neutron
    file_name = 'data/xs_macro.xml'
    source_material_id = '7'
    block = concrete
  []
  [Water]
    type = FileTransportMaterial
    transport_system = Neutron
    file_name = 'data/water_macro_xs.xml'
    source_material_id = '1'
    block = light_water
  []
[]

[Executioner]
  type = Steady
  solve_type = PJFNK
  petsc_options_iname = '-ksp_gmres_restart'
  petsc_options_value = ' 600'

  l_max_its = 50
  nl_abs_tol = 1e-12
  #nl_rel_tol = 1e-13

  automatic_scaling = true
  off_diagonals_in_auto_scaling = true
  compute_scaling_once = false
[]

[Outputs]
  exodus = true
[]

(examples/2D_subcritical_reactor_1_group/neutronics_transport_eigen.i)

[Mesh]
  [NeutronicsDomain]
    type = FileMeshGenerator
    file = Radiation_Mesh.e
  []
[]

[TransportSystems]
  [Neutron]
    num_groups = 1
    max_anisotropy = 0
    scheme = saaf_cfem
    particle_type = neutron

    flux_moment_names = 'Flux_Moment'
    debug_disable_fission = false
    eigen = true
    constant_ic = '1.0'

    order = FIRST
    family = LAGRANGE

    n_azimuthal = 5
    n_polar = 5

    vacuum_boundaries = 'vacuum'
  []
[]

[TransportMaterials]
  [Air]
    type = FileTransportMaterial
    transport_system = Neutron
    file_name = 'macro_xs.xml'
    source_material_id = '5'
    block = '2 7 8 9 ' #
  []

  [Cladding]
    type = FileTransportMaterial
    transport_system = Neutron
    file_name = 'macro_xs.xml'
    source_material_id = '7' #
    block = 4
  []

  [Fuel]
    type = FileTransportMaterial
    transport_system = Neutron
    file_name = 'macro_xs.xml'
    source_material_id = '6' #
    block = 3
  []

  [Graphite]
    type = FileTransportMaterial
    transport_system = Neutron
    file_name = 'macro_xs.xml'
    source_material_id = '8' #
    block = '10 11'
  []

  [Control]
    type = FileTransportMaterial
    transport_system = Neutron
    file_name = 'macro_xs.xml'
    source_material_id = '9' #
    block = 6
  []
[]

[Executioner]
  type = Eigenvalue

  initial_eigenvalue = 1.2

  solve_type = PJFNK
  petsc_options_iname = '-ksp_gmres_restart'
  petsc_options_value = ' 600'

  nl_abs_tol = 1e-12

  automatic_scaling = true
  off_diagonals_in_auto_scaling = true
  compute_scaling_once = false
[]

[Outputs]
  exodus = true
  execute_on = 'TIMESTEP_END'
[]

(examples/2D_diff_water_steady_cgfem/test.i)

# A simple test case with a purely absorbing medium and a point source in the
# middle of the domain.

[Mesh]
  [domain]
    type = CartesianMeshGenerator
    dim = 2
    dx = '3.5 3 3.5'
    dy = '3.5 3 3.5'
    ix = '35 31 35'
    iy = '35 31 35'
    subdomain_id = '
      2 2 2
      2 1 2
      2 2 2'
  []
[]

[TransportSystems]
  [Neutron]
    num_groups = 2
    scheme = diffusion_cfem
    particle_type = neutron

    order = FIRST
    family = LAGRANGE

    vacuum_boundaries = 'left right top bottom'

    point_source_locations = '5.0 5.0 0.0'
    point_source_moments = '1e3 0.0'
    point_source_anisotropies = '0'
  []
[]

[TransportMaterials]
  [Water]
    type = FileTransportMaterial
    transport_system = Neutron
    file_name = 'macro_xs.xml'
    source_material_id = '1'
    block = '1 2'
  []
[]

[Executioner]
  type = Steady
  solve_type = PJFNK

  l_max_its = 50
  nl_rel_tol = 1e-12
[]

[Outputs]
  exodus = true
[]

(examples/2D_subcritical_1_group_cgfem/neutronics_uncollided.i)

[Mesh]
  [NeutronicsDomain]
    type = FileMeshGenerator
    file = Storage_Radiation_Mesh_Tris.e
  []
  uniform_refine = 0
[]

[UncollidedFlux]
  [Neutron]
    uncollided_flux_treatment = ray-tracing
    num_groups = 1
    max_anisotropy = 1

    volumetric_source_blocks = fuel
    volumetric_source_moments = '40.5717'
    volumetric_source_anisotropies = '0'

    rt_n_polar = 4
  []
[]

[TransportMaterials]
  [Air]
    type = FileTransportMaterial
    transport_system = 'Neutron'
    file_name = 'data/macro_xs.xml'
    source_material_id = '5'
    block = air
  []
  [Cladding]
    type = FileTransportMaterial
    transport_system = 'Neutron'
    file_name = 'data/macro_xs.xml'
    source_material_id = '6'
    block = cladding
  []
  [Fuel]
    type = FileTransportMaterial
    transport_system = 'Neutron'
    file_name = 'data/macro_xs.xml'
    source_material_id = '7'
    block = fuel
  []
  [Wood]
    type = FileTransportMaterial
    transport_system = 'Neutron'
    file_name = 'data/macro_xs.xml'
    source_material_id = '8'
    block = wood
  []
[]

[Problem]
  type = FEProblem
  solve = false
  kernel_coverage_check = false
  material_coverage_check = false
[]

[Executioner]
  type = Steady
[]

[Outputs]
  exodus = true
  execute_on = 'TIMESTEP_END'
[]

(examples/2D_C5G7_full/neutronics.i)

# Solving the 2D C5G7 light water reactor benchmark problem.

[Mesh]
  [NeutronicsDomain]
    type = FileMeshGenerator
    file = 2D_C5G7_Full.e
  []
  uniform_refine = 0
[]

[TransportSystems]
  [Neutron]
    num_groups = 7
    max_anisotropy = 0
    scheme = saaf_cfem
    particle_type = neutron

    flux_moment_names = 'Flux_Moment'
    debug_disable_fission = false
    eigen = true
    constant_ic = '1.0'

    order = FIRST
    family = LAGRANGE

    n_azimuthal = 3
    n_polar = 3

    reflective_boundaries = 'reflective'
    vacuum_boundaries = 'vacuum'
  []
[]

[TransportMaterials]
  [UO2]
    type = FileTransportMaterial
    transport_system = Neutron
    file_name = 'C5G7_XS.xml'
    source_material_id = '0'
    block = 'uo2'
  []

  [MOX_43]
    type = FileTransportMaterial
    transport_system = Neutron
    file_name = 'C5G7_XS.xml'
    source_material_id = '1'
    block = 'mox43'
  []

  [MOX_70]
    type = FileTransportMaterial
    transport_system = Neutron
    file_name = 'C5G7_XS.xml'
    source_material_id = '2'
    block = 'mox70'
  []

  [MOX_87]
    type = FileTransportMaterial
    transport_system = Neutron
    file_name = 'C5G7_XS.xml'
    source_material_id = '3'
    block = 'mox87'
  []

  [Fission_Chambers]
    type = FileTransportMaterial
    transport_system = Neutron
    file_name = 'C5G7_XS.xml'
    source_material_id = '4'
    block = 'fission_chambers'
  []

  [Guide_Tubes]
    type = FileTransportMaterial
    transport_system = Neutron
    file_name = 'C5G7_XS.xml'
    source_material_id = '5'
    block = 'guide_tubes'
  []

  [Moderator_Reflector]
    type = FileTransportMaterial
    transport_system = Neutron
    file_name = 'C5G7_XS.xml'
    source_material_id = '6'
    block = 'moderator reflector'
  []
[]

[Postprocessors]
  [TotalFlux]
    type = TotalFluxPostProcessor
    num_groups = 7
    group_scalar_fluxes = 'Flux_Moment_1_0_0 Flux_Moment_2_0_0 Flux_Moment_3_0_0 Flux_Moment_4_0_0 Flux_Moment_5_0_0 Flux_Moment_6_0_0 Flux_Moment_7_0_0'
    execute_on = LINEAR
  []
[]

[VectorPostprocessors]
  [EigenValues]
    type = Eigenvalues
    inverse_eigenvalue = true
  []
[]

[Executioner]
  type = Eigenvalue

  initial_eigenvalue = 1.0
  normal_factor = 1.0
  normalization = 'TotalFlux'
  free_power_iterations = 6

  solve_type = PJFNK
  #petsc_options_iname = '-pc_type -pc_hypre_type -ksp_gmres_restart'
  #petsc_options_value = ' hypre    boomeramg      600'
  petsc_options_iname = '-ksp_gmres_restart'
  petsc_options_value = ' 600'

  nl_abs_tol = 1e-8

  automatic_scaling = true
  off_diagonals_in_auto_scaling = true
  compute_scaling_once = false
[]

[Outputs]
  csv = true
  exodus = true
  execute_on = 'TIMESTEP_END'
[]

(examples/2D_subcritical_reactor_4_groups/neutronics_uncollided.i)

# An input deck which compute the uncollided flux in the assembly with the SASF approach.
# The SASF method in 2D assumes a plane-wave point source which is not physically realistic for most
# particle sources.

[Mesh]
  [NeutronicsDomain]
    type = FileMeshGenerator
    file = 2D_quarter_core_nsr.e
  []
  uniform_refine = 0
[]

[UncollidedFlux]
  [Neutron]
    uncollided_flux_treatment = sasf
    num_groups = 4
    max_anisotropy = 0

    point_source_locations = '0.0 0.0 0.0'
    point_source_moments = '1.5'
    point_source_anisotropies = '0'

    sasf_near_source_boundary = 'near_source'
    sasf_vacuum_boundaries = 'vacuum'
    sasf_near_source_cross_sections = '9.683916522303286e-05 0.00023453400588691886 0.00043185580016127494 0.0005108576781703109'

    is_conservative_transfer_src = false
  []
[]

[TransportMaterials]
  [Air]
    type = FileTransportMaterial
    transport_system = Neutron
    file_name = 'macro_xs.xml'
    source_material_id = '5' #
    block = 'air_gap'
  []

  [Cladding]
    type = FileTransportMaterial
    transport_system = Neutron
    file_name = 'macro_xs.xml'
    source_material_id = '7' #
    block = 'cladding'
  []

  [Fuel]
    type = FileTransportMaterial
    transport_system = Neutron
    file_name = 'macro_xs.xml'
    source_material_id = '6' #
    block = 'fuel'
  []

  [Graphite]
    type = FileTransportMaterial
    transport_system = Neutron
    file_name = 'macro_xs.xml'
    source_material_id = '8' #
    block = 'core_graphite reflector_graphite'
  []

  [Control]
    type = FileTransportMaterial
    transport_system = Neutron
    file_name = 'macro_xs.xml'
    source_material_id = '8' #
    block = 'control_or_graphite'
  []
[]

[Executioner]
  type = Steady
  solve_type = PJFNK
  petsc_options_iname = '-ksp_gmres_restart'
  petsc_options_value = ' 600'

  l_max_its = 50
  nl_abs_tol = 1e-12

  automatic_scaling = true
  off_diagonals_in_auto_scaling = true
  compute_scaling_once = false
[]

(examples/2D_C5G7_simplified/neutronics.i)

[Mesh]
  [NeutronicsDomain]
    type = FileMeshGenerator
    file = 2D_C5G7_simplified.e
  []
  uniform_refine = 0
[]

[TransportSystems]
  [Neutron]
    num_groups = 7
    max_anisotropy = 0
    scheme = saaf_cfem
    particle_type = neutron

    flux_moment_names = 'Flux_Moment'
    debug_disable_fission = false
    eigen = true
    constant_ic = '1.0'

    order = FIRST
    family = LAGRANGE

    n_azimuthal = 5
    n_polar = 5

    reflective_boundaries = 'reflective'
  []
[]

[TransportMaterials]
  [FuelClad]
    type = FileTransportMaterial
    transport_system = Neutron
    file_name = 'C5G7_XS.xml'
    source_material_id = '0'
    block = 'fuel'
  []

  [Moderator]
    type = FileTransportMaterial
    transport_system = Neutron
    file_name = 'C5G7_XS.xml'
    source_material_id = '6'
    block = 'moderator'
  []

  [Void]
    type = FileTransportMaterial
    transport_system = Neutron
    file_name = 'C5G7_XS.xml'
    source_material_id = '7'
    block = 'air'
  []
[]

[Postprocessors]
  [TotalFlux]
    type = TotalFluxPostProcessor
    num_groups = 7
    group_scalar_fluxes = 'Flux_Moment_1_0_0 Flux_Moment_2_0_0 Flux_Moment_3_0_0 Flux_Moment_4_0_0 Flux_Moment_5_0_0 Flux_Moment_6_0_0 Flux_Moment_7_0_0'
    execute_on = LINEAR
  []
[]

[VectorPostprocessors]
  [EigenValues]
    type = Eigenvalues
    inverse_eigenvalue = true
  []
[]

[Executioner]
  type = Eigenvalue

  initial_eigenvalue = 1.0
  normal_factor = 1.0
  normalization = 'TotalFlux'
  free_power_iterations = 6

  solve_type = PJFNK
  #petsc_options_iname = '-pc_type -pc_hypre_type -ksp_gmres_restart'
  #petsc_options_value = ' hypre    boomeramg      600'
  petsc_options_iname = '-ksp_gmres_restart'
  petsc_options_value = ' 600'

  nl_abs_tol = 1e-8

  automatic_scaling = true
  off_diagonals_in_auto_scaling = true
  compute_scaling_once = false
[]

[Outputs]
  csv = true
  exodus = true
  execute_on = 'TIMESTEP_END'
[]

(examples/2D_subcritical_reactor_4_groups/neutronics_eigen.i)

[Mesh]
  [NeutronicsDomain]
    type = FileMeshGenerator
    file = 2D_quarter_core.e
  []
  uniform_refine = 0
[]

[TransportSystems]
  [Neutron]
    num_groups = 4
    max_anisotropy = 0
    scheme = saaf_cfem
    particle_type = neutron
    flux_moment_names = 'Flux_Moment'
    debug_disable_fission = false
    eigen = true

    order = FIRST
    family = LAGRANGE

    constant_ic = '1.0'

    n_azimuthal = 3
    n_polar = 3

    vacuum_boundaries = 'vacuum'
    reflective_boundaries = 'reflective'
  []
[]

[TransportMaterials]
  [Air]
    type = FileTransportMaterial
    transport_system = Neutron
    file_name = 'macro_xs.xml'
    source_material_id = '5' #
    block = 'air_gap src_air'
  []

  [Cladding]
    type = FileTransportMaterial
    transport_system = Neutron
    file_name = 'macro_xs.xml'
    source_material_id = '7' #
    block = 'cladding'
  []

  [Fuel]
    type = FileTransportMaterial
    transport_system = Neutron
    file_name = 'macro_xs.xml'
    source_material_id = '6' #
    block = 'fuel'
  []

  [Graphite]
    type = FileTransportMaterial
    transport_system = Neutron
    file_name = 'macro_xs.xml'
    source_material_id = '8' #
    block = 'core_graphite reflector_graphite'
  []

  [Control]
    type = FileTransportMaterial
    transport_system = Neutron
    file_name = 'macro_xs.xml'
    source_material_id = '8' #
    block = 'control_or_graphite'
  []
[]

[Executioner]
  type = Eigenvalue

  initial_eigenvalue = 1.2
  free_power_iterations = 6

  solve_type = PJFNK
  petsc_options_iname = '-pc_type -pc_hypre_type -ksp_gmres_restart'
  petsc_options_value = ' hypre    boomeramg      300'

  nl_abs_tol = 1e-12

  automatic_scaling = true
  off_diagonals_in_auto_scaling = true
  compute_scaling_once = false
[]

[VectorPostprocessors]
  [EigenValues]
    type = Eigenvalues
    inverse_eigenvalue = true
  []
[]

[Outputs]
  csv = true
  exodus = true
  execute_on = 'TIMESTEP_END'
[]

Overview
Example Input File Syntax
Input Parameters
Input Files