mcdc.Tally#

Quantities measured during the simulation.

Parameters:

name (str, optional) – User-facing tally name.
scores (list of str, optional) – Scores to accumulate. Track-length scores are "flux", "density", "collision", "capture", and "fission"; surface-crossing scores are "current-net", "current-in", and "current-out"; the collision score is "energy_deposition". Scores from different estimator families cannot be mixed.
surface (Surface, optional) – Surface filter. Required for a surface-crossing tally unless cell is provided.
cell (Cell, optional) – Cell filter.
mesh (MeshBase, optional) – Spatial mesh filter for track-length or collision tallies.
mu (sequence of float, optional) – Polar-cosine bin boundaries.
azi (sequence of float, optional) – Azimuthal-angle bin boundaries in radians.
polar_reference (sequence of 3 float, optional) – Reference direction for the angular filters.
energy (sequence of float or "all_groups", optional) – Energy-bin boundaries in eV, or "all_groups" in multigroup mode.
time (sequence of float, optional) – Time-bin boundaries in seconds.

Returns:

Concrete tally selected from scores.

Return type:

TallySurfaceCrossing, TallyTracklength, or TallyCollision

Examples

Score flux and fission on a structured mesh:

>>> import numpy as np
>>> import mcdc
>>> mesh = mcdc.MeshStructured(z=np.linspace(0.0, 10.0, 101))
>>> tally = mcdc.Tally(
...     name="Axial flux",
...     mesh=mesh,
...     scores=["flux", "fission"],
...     energy=[0.0, 1.0e6, 20.0e6],
... )

Score net current crossing a surface:

>>> boundary = mcdc.Surface.PlaneZ(z=10.0)
>>> current = mcdc.Tally(surface=boundary, scores=["current-net"])

Filter a track-length tally by cell, angle, and time:

>>> material = mcdc.MaterialMG(capture=np.array([1.0]))
>>> lower = mcdc.Surface.PlaneZ(z=0.0)
>>> upper = mcdc.Surface.PlaneZ(z=10.0)
>>> cell = mcdc.Cell(region=+lower & -upper, fill=material)
>>> filtered_flux = mcdc.Tally(
...     cell=cell,
...     scores=["flux", "capture"],
...     mu=np.linspace(-1.0, 1.0, 11),
...     azi=np.linspace(-np.pi, np.pi, 17),
...     time=[0.0, 1.0e-6, 2.0e-6],
... )

Compile the geometry to resolve the cell’s bounding surfaces:

>>> model = mcdc.Simulation(name="Cell-current example")
>>> model.set_model([cell])
>>> model.compile()

Score current entering and leaving the cell through any of its boundaries:

>>> cell_current = mcdc.Tally(
...     cell=cell,
...     scores=["current-net", "current-in", "current-out"],
... )

A cell-only surface-crossing filter scores genuine changes in cell membership. Crossings of surfaces inside the cell that do not enter or leave the cell are not scored.

Restrict the cell current to one particular boundary surface:

>>> upper_surface_current = mcdc.Tally(
...     surface=upper,
...     cell=cell,
...     scores=["current-net", "current-in", "current-out"],
... )
>>> model.set_tallies([cell_current, upper_surface_current])

With both filters, the surface selects where crossings are scored, while the cell determines whether each crossing is incoming or outgoing.

Score energy deposition with both cell and mesh filters:

>>> deposition = mcdc.Tally(
...     cell=cell,
...     mesh=mesh,
...     scores=["energy_deposition"],
... )

Use one energy bin per multigroup energy group:

>>> group_flux = mcdc.Tally(
...     cell=cell,
...     scores=["flux"],
...     energy="all_groups",
... )