Monte Carlo method for array criticality calculations

Journal Article · Sun Aug 01 04:00:00 EDT 1976 · Nucl. Technol.; (United States)

OSTI ID:7354986

Dickinson, D ^[1]; Whitesides, G E

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1. Rockwell International, Golden, CO

The Monte Carlo method for solving neutron transport problems consists of mathematically tracing paths of individual neutrons collision by collision until they are lost by absorption or leakage. The fate of the neutron after each collision is determined by the probability distribution functions that are formed from the neutron cross-section data. These distributions are sampled statistically to establish the successive steps in the neutron's path. The resulting data, accumulated from following a large number of batches, are analyzed to give estimates of k/sub eff/ and other collision-related quantities. The use of electronic computers to produce the simulated neutron histories, initiated at Los Alamos Scientific Laboratory, made the use of the Monte Carlo method practical for many applications. In analog Monte Carlo simulation, the calculation follows the physical events of neutron scattering, absorption, and leakage. To increase calculational efficiency, modifications such as the use of statistical weights are introduced. The Monte Carlo method permits the use of a three-dimensional geometry description and a detailed cross-section representation. Some of the problems in using the method are the selection of the spatial distribution for the initial batch, the preparation of the geometry description for complex units, and the calculation of error estimates for region-dependent quantities such as fluxes. The Monte Carlo method is especially appropriate for criticality safety calculations since it permits an accurate representation of interacting units of fissile material. Dissimilar units, units of complex shape, moderators between units, and reflected arrays may be calculated. Monte Carlo results must be correlated with relevant experimental data, and caution must be used to ensure that a representative set of neutron histories is produced.

• OSTI ID: 7354986
• Journal Information: Nucl. Technol.; (United States), Journal Name: Nucl. Technol.; (United States) Vol. 30:2; ISSN NUTYB
• Country of Publication: United States
• Language: English

Related Subjects

42 ENGINEERING
420203* -- Engineering-- Handling Equipment & Procedures
COMPUTER CALCULATIONS
CRITICALITY
FISSILE MATERIALS
FISSIONABLE MATERIALS
MONTE CARLO METHOD
NEUTRAL-PARTICLE TRANSPORT
NEUTRON TRANSPORT
RADIATION TRANSPORT
SAFETY