APOLLO: a general code for transport, slowing-down and thermalization calculations in heterogeneous media

Conference · Mon Jan 01 04:00:00 EST 1973

OSTI ID:4392112

Kavenoky, A

From national topical meeting on mathematical models and computational techniques for analysis of nuclear systems; Ann Arbor, Michigan, USA (8 Apr 1973). In mathematical models and computational techniques for analysis of nuclear systems. APOLLO calculates the space-and-energy-dependent flux for a one dimensional medium, in the multigroup approximation of the transport equation. For a one dimensional medium, refined collision probabilities have been developed for the resolution of the integral form of the transport equation; these collision probabilities increase accuracy and save computing time. The interaction between a few cells can also be treated by the multicell option of APOLLO. The diffusion coefficient and the material buckling can be computed in the various B and P approximations with a linearly anisotropic scattering law, even in the thermal range of the spectrum. Eventually this coefficient is corrected for streaming by use of Benoist's theory. The self-shielding of the heavy isotopes is treated by a new and accurate technique which preserves the reaction rates of the fundamental fine structure flux. APOLLO can perform a depletion calculation for one cell, a group of cells or a complete reactor. The results of an APOLLO calculation are the space-and-energy-dependent flux, the material buckling or any reaction rate; these results can also be macroscopic cross sections used as input data for a 2D or 3D depletion and diffusion code in reactor geometry. 10 references. (auth)

• Research Organization: Centre d'Etudes Nucleaires, Saclay, France; American Nuclear Society. Michigan Section
• NSA Number: NSA-29-007172
• OSTI ID: 4392112
• Report Number(s): CONF-730414--P1
• Country of Publication: Country unknown/Code not available
• Language: English

Related Subjects

*REACTOR KINETICS-- NEUTRON TRANSPORT THEORY
A CODES
ENERGY DEPENDENCE
MULTIGROUP THEORY
N79100* --Reactors--Reactor Theory & Calculation
NEUTRON FLUX
ONE-DIMENSIONAL CALCULATIONS
REACTION KINETICS
SPACE DEPENDENCE SPACE DEPENDENCE