MatMCNP: A Code for Producing Material Cards for MCNP
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.
- Authors:
-
- Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)
- American Structurepoint, Inc., Indianapolis, IN (United States)
- Publication Date:
- Research Org.:
- Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)
- Sponsoring Org.:
- USDOE National Nuclear Security Administration (NNSA)
- OSTI Identifier:
- 1323135
- Report Number(s):
- SAND2014-17693
537405; TRN: US1601898
- DOE Contract Number:
- AC04-94AL85000
- Resource Type:
- Technical Report
- Country of Publication:
- United States
- Language:
- English
- Subject:
- 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
Citation Formats
DePriest, Kendall Russell, and Saavedra, Karen C. MatMCNP: A Code for Producing Material Cards for MCNP. United States: N. p., 2014.
Web. doi:10.2172/1323135.
DePriest, Kendall Russell, & Saavedra, Karen C. MatMCNP: A Code for Producing Material Cards for MCNP. United States. https://doi.org/10.2172/1323135
DePriest, Kendall Russell, and Saavedra, Karen C. 2014.
"MatMCNP: A Code for Producing Material Cards for MCNP". United States. https://doi.org/10.2172/1323135. https://www.osti.gov/servlets/purl/1323135.
@article{osti_1323135,
title = {MatMCNP: A Code for Producing Material Cards for MCNP},
author = {DePriest, Kendall Russell and Saavedra, Karen C.},
abstractNote = {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.},
doi = {10.2172/1323135},
url = {https://www.osti.gov/biblio/1323135},
journal = {},
number = ,
volume = ,
place = {United States},
year = {Mon Sep 01 00:00:00 EDT 2014},
month = {Mon Sep 01 00:00:00 EDT 2014}
}
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