AECL

IAEA TECDOC-887

MCNP

MCNPX

Abstract
               The MCNPX (Monte Carlo N-Particle Transport Code System) code has been used for modeling and simulation of a half core of CANDU (CANada Deuterium-Uranium) reactor, both homogenous and heterogeneous model for the reactor core are designed. The fuel is burnt in normal operation conditions of CANDU reactors. Natural uranium fuel is used in the model. The multiplication factor for homogeneous and heterogeneous reactor core is calculated and compared during fuel burnup. The concentration of both uranium and plutonium isotopes are analysed in the model. The flux and power distributions through channels are calculated.</jats:p

A. EL-Khawlani

M. Aziz

M. Ismail

A. Y. Ellithi

Crossref

Kerntechnik

Analysis of a homogenous and heterogeneous stylized half core of a CANDU reactor

Analysis of a homogenous and heterogeneous stylized half core of a CANDU reactor

Abstract

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