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DOE OSTI · 1903539

MCNP Intrinsic Source Constructor (MISC): A User’s Guide

Solomon Jr., Clell (C. J.) Jeffrey·Bates, Cameron Russell·Marcath, Matthew James
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Abstract

The MCNP Intrinsic Source Constructor (MISC) facilitates the construction of MCNP source definition cards for radioactive materials. Given a material nuclide specification, either weight or atom fractions of those nuclides, and either a weight or atom density of the material, MISC will construct the source spectrum distribution of a specified particle type. MISC also calculates the absolute number of particles emitted per second per unit mass or volume. Furthermore, if decay chain information is available in the data sets selected by the user, the initial isotopes specified by the user can be approximately decayed to the resulting isotopes after a specified amount of time.

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73 NUCLEAR PHYSICS AND RADIATION PHYSICS97 MATHEMATICS AND COMPUTING

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BibTeXRIS

Solomon Jr., Clell (C. J.) Jeffrey, Bates, Cameron Russell, Marcath, Matthew James. 2021-04-24. MCNP Intrinsic Source Constructor (MISC): A User’s Guide. https://doi.org/10.2172/1903539

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