A software tool to create material descriptions for neutron transport calculations

What is it about?

The material descriptions required for neutron transport calculations can be time consuming, and they are frequently prepared incorrectly. The MatMCNP software takes basic information on a material provided by a user and produces a material description that can be used directly by the popular MCNP code. The descriptions are also easily translated to the format required by other neutron transport tools.

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Why is it important?

MatMCNP allows for material specification as either atomic or weight percent (fractions). The software 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 energy-dependent neutron and gamma fluences into dose in the material specified.