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Title: LANL Summer 2016 Report

The Monte Carlo N-Particle (MCNP) transport code developed at Los Alamos National Laboratory (LANL) utilizes nuclear cross-section data in a compact ENDF (ACE) format. The accuracy of MCNP calculations depends on the accuracy of nuclear ACE data tables, which depends on the accuracy of the original ENDF files. There are some noticeable differences in ENDF files from one generation to the next, even among the more common fissile materials. As the next generation of ENDF files is being prepared, several software tools were developed to simulate a large number of benchmarks in MCNP (over 1000), collect data from these simulations, and visually represent the results.
Authors:
 [1]
  1. Los Alamos National Lab. (LANL), Los Alamos, NM (United States)
Publication Date:
OSTI Identifier:
1312622
Report Number(s):
LA-UR--16-26595
TRN: US1601821
DOE Contract Number:
AC52-06NA25396
Resource Type:
Technical Report
Research Org:
Los Alamos National Lab. (LANL), Los Alamos, NM (United States)
Sponsoring Org:
USDOE
Country of Publication:
United States
Language:
English
Subject:
73 NUCLEAR PHYSICS AND RADIATION PHYSICS; 97 MATHEMATICS AND COMPUTING; NUCLEAR DATA COLLECTIONS; FISSILE MATERIALS; ACCURACY; MONTE CARLO METHOD; DATA; M CODES; CROSS SECTIONS; BENCHMARKS; COMPUTERIZED SIMULATION; NEUTRON TRANSPORT; DATA VISUALIZATION