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Title: MatMCNP: A Code for Producing Material Cards for MCNP

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:
 [1] ;  [2]
  1. Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)
  2. American Structurepoint, Inc., Indianapolis, IN (United States)
Publication Date:
OSTI Identifier:
1323135
Report Number(s):
SAND2014--17693
537405; TRN: US1601898
DOE Contract Number:
AC04-94AL85000
Resource Type:
Technical Report
Research Org:
Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States)
Sponsoring Org:
USDOE National Nuclear Security Administration (NNSA)
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