MCNP-DSP Calculations of the ²⁵²Cf-Source-Driven Noise Analysis Measurements of Highly Enriched Uranium Metal Cylinders

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

OSTI ID:96845

Valentine, T. E. ^[1]; Mihalczo, J. T. ^[1]

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1. Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Instrumentation and Controls Division

This paper presents calculations of the ²⁵²Cf-source-driven noise analysis measurements for subcritical highly enriched uranium metal cylinders using the Monte Carlo code MCNP-DSP. This code directly calculates the noise analysis data from the ²⁵²Cf-source-driven noise analysis method for both neutron and gamma ray detectors. Direct calculation of experimental observables by the Monte Carlo method allows for the benchmarking of the calculational model and the cross sections and for determining the bias in the calculation.

• Research Organization: Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States). Instrumentation and Controls Division
• Sponsoring Organization: USDOE National Nuclear Security Administration (NNSA), Nuclear Criticality Safety Program (NCSP)
• DOE Contract Number: AC05-84OR21400
• OSTI ID: 96845
• Report Number(s): CONF-9509100-3; ON: DE95014579; TRN: 95:018502
• Resource Relation: Conference: ICNC '95: 5. International Conference on Nuclear Criticality Safety, Albuquerque, NM (United States), 17-22 Sep 1995; Other Information: PBD: [1995]
• Country of Publication: United States
• Language: English

Related Subjects

73 NUCLEAR PHYSICS AND RADIATION PHYSICS
97 MATHEMATICS AND COMPUTING
NEUTRON TRANSPORT
M CODES
HIGHLY ENRICHED URANIUM
SPECTRAL DENSITY
MONTE CARLO METHOD
CALIFORNIUM 252
BENCHMARKS
NUCLEAR DATA COLLECTIONS
MULTIPLICATION FACTORS
PROMPT NEUTRONS
CROSS SECTIONS
CRITICALITY
Nuclear Criticality Safety Program (NCSP)
Monte Carlo Code
MCNP-DSP
Neutron and Gamma Ray Detectors
Evaluated Nuclear Data File (ENDF)
Autopower Spectral Densities (APSDs)
Fissile Material
Cross-Power Spectral Densities (CPSDs)