Validation of MCNP6 Version 1.0 with the ENDF/B-VII.1 Cross Section Library for Uranium Metal, Oxide, and Solution Systems on the High Performance Computing Platform Moonlight

Technical Report · Thu Apr 21 00:00:00 EDT 2016

DOI:https://doi.org/10.2172/1334657· OSTI ID:1334657

Chapman, Bryan Scott ^[1]; MacQuigg, Michael Robert ^[1]; Wysong, Andrew Russell ^[1]

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1. Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

In this document, the code MCNP is validated with ENDF/B-VII.1 cross section data under the purview of ANSI/ANS-8.24-2007, for use with uranium systems. MCNP is a computer code based on Monte Carlo transport methods. While MCNP has wide reading capability in nuclear transport simulation, this validation is limited to the functionality related to neutron transport and calculation of criticality parameters such as k_eff.

• Research Organization: Los Alamos National Laboratory (LANL), Los Alamos, NM (United States)
• Sponsoring Organization: USDOE National Nuclear Security Administration (NNSA)
• DOE Contract Number: AC52-06NA25396
• OSTI ID: 1334657
• Report Number(s): LA-UR-16-29185; TRN: US1700811
• Country of Publication: United States
• Language: English

Related Subjects

97 MATHEMATICS AND COMPUTING
73 NUCLEAR PHYSICS AND RADIATION PHYSICS
URANIUM
URANIUM OXIDES
URANYL FLUORIDES
URANYL NITRATES
SOLUTIONS
NUCLEAR DATA COLLECTIONS
VALIDATION
CROSS SECTIONS
NEUTRON TRANSPORT
MONTE CARLO METHOD
M CODES
CRITICALITY
COMPUTERIZED SIMULATION
URANIUM 235
Criticality Safety
MCNP