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DOE OSTI · 1323135

MatMCNP: A Code for Producing Material Cards for MCNP

DePriest, Kendall Russell·Saavedra, Karen C.
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Abstract

A code for generating MCNP material cards (MatMCNP) has been written and verified for naturally occurring, stable isotopes. The program allows for material specification as either atomic or weight percent (fractions). MatMCNP also permits the specification of enriched lithium, boron, and/or uranium. In addition to producing the material cards for MCNP, the code calculates the atomic (or number) density in atoms/barn-cm as well as the multiplier that should be used to convert neutron and gamma fluences into dose in the material specified.

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73 NUCLEAR PHYSICS AND RADIATION PHYSICS97 MATHEMATICS AND COMPUTINGBORONDENSITYGAMMA RADIATIONLITHIUMM CODESNEUTRAL-PARTICLE TRANSPORTNEUTRONSRADIATION DOSESSTABLE ISOTOPESURANIUM

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DePriest, Kendall Russell, Saavedra, Karen C.. 2014-09-01. MatMCNP: A Code for Producing Material Cards for MCNP. https://doi.org/10.2172/1323135

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