Void / Vacuum

Gases & Other

Density: 0 g/cm³

Vacuum or void region. No material is present; particles stream freely through these regions.

Composition

Nuclide / Element	Fraction

Code output

MCNP

c Void / Vacuum
c No material card needed.
c Use material number 0 on the cell card:
c   1  0  -1    $ void cell inside surface 1

SCONE reads ACE libraries: build nuclearData.materials entries from these nuclides using ZAID plus a temperature/table suffix that exists in your aceLibrary, with number densities in atoms/barn-cm. See the SCONE nuclear data tutorial and the NRDP library reference.

Notes

In MCNP, void cells use material number 0 with no density. In Serpent, use the keyword "void". In OpenMC, leave Cell.fill as None.

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