MatMCNP: A Code for Producing Material Cards for MCNP

Technical Report · Mon Sep 01 00:00:00 EDT 2014

DOI:https://doi.org/10.2172/1323135· OSTI ID:1323135

DePriest, Kendall Russell ^[1]; Saavedra, Karen C. ^[2]

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1. Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)
2. American Structurepoint, Inc., Indianapolis, IN (United States)

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.

• Research Organization: Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)
• Sponsoring Organization: USDOE National Nuclear Security Administration (NNSA)
• DOE Contract Number: AC04-94AL85000
• OSTI ID: 1323135
• Report Number(s): SAND2014-17693; 537405; TRN: US1601898
• Country of Publication: United States
• Language: English

Related Subjects

97 MATHEMATICS AND COMPUTING
73 NUCLEAR PHYSICS AND RADIATION PHYSICS
M CODES
STABLE ISOTOPES
NEUTRONS
LITHIUM
URANIUM
BORON
DENSITY
RADIATION DOSES
GAMMA RADIATION
NEUTRAL-PARTICLE TRANSPORT