Open AccessMCNP: Neutron benchmark problems
Author(s) -
Daniel J. Whalen,
Devin Cardon,
J.L. Uhle,
J.S. Hendricks
Publication year - 1991
Language(s) - English
Resource type - Reports
DOI - 10.2172/10103487
Subject(s) - criticality , radiation transport , benchmark (surveying) , neutron transport , monte carlo method , neutron , nuclear engineering , nuclear data , code (set theory) , physics , computer science , electromagnetic shielding , nuclear physics , statistical physics , computational physics , mathematics , engineering , statistics , geology , geodesy , set (abstract data type) , quantum mechanics , programming language
The recent widespread and increased use of radiation transport codes has produced greater user and institutional demand for assurances that such codes give correct results. Responding to these requirements for code validation, the general purpose Monte Carlo transport code MCNP has been tested on criticality, pulsed sphere, and shielding neutron problem families. Results for each were compared to experimental data. MCNP successfully predicted the experimental results of all three families within the expected data and statistical uncertainties. These successful predictions demonstrate that MCNP can successfully model a broad spectrum of neutron transport problems. 18 refs., 27 figs., 4 tabs.
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