MCNP Calculations for Criticality Safety Benchmarks with ENDF/B-V and ENDF/B-VI Libraries

Conference · Sun Sep 17 00:00:00 EDT 1995

OSTI ID:97187

Iverson, John L. ^[1]; Mosteller, Russell D. ^[2]

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1. Univ. of Arizona, Tucson, AZ (United States). Department of Nuclear Engineering
2. Los Alamos National Laboratory (LANL), Los Alamos, NM (United States). Reactor Design and Analysis Group, Technology and Safety Assessment Division

The MCNP Monte Carlo code, in conjunction with its continuous-energy ENDF/B-V and ENDF/B-VI cross-section libraries, has been benchmarked against results from 27 different critical experiments. The predicted values of k_eff are in excellent agreement with the benchmarks, except for the ENDF/B-V results for solutions of plutonium nitrate and, to a lesser degree, for the ENDF/B-V and ENDF/B-VI results for a bare sphere of ²³³U.

• Research Organization: Los Alamos National Laboratory (LANL), Los Alamos, NM (United States). Reactor Design and Analysis Group, Technology and Safety Assessment Division; Univ. of Arizona, Tucson, AZ (United States). Department of Nuclear Engineering
• Sponsoring Organization: USDOE National Nuclear Security Administration (NNSA), Nuclear Criticality Safety Program (NCSP)
• DOE Contract Number: W-7405-ENG-36
• OSTI ID: 97187
• Report Number(s): LA-UR--95-1856; CONF-9509100--25; ON: DE95015255
• Country of Publication: United States
• Language: English

Related Subjects

73 NUCLEAR PHYSICS AND RADIATION PHYSICS
97 MATHEMATICS AND COMPUTING
BENCHMARKS
CRITICALITY
CROSS SECTIONS
Cross-Section Libraries
Evaluated Nuclear Data File (ENDF)
Fast-Spectrum Plutonium Systems
Highly Enriched Uranium (HEU)
Jemima Experiments
M CODES
MONTE CARLO METHOD
Monte Carlo Code
NUCLEAR DATA COLLECTIONS
Nuclear Criticality Safety Program (NCSP)
REACTOR SAFETY
URANIUM 233