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a b s t r a c tThe analysis of neutron kinetics relies on the knowledge of adjoint-weighted kinetics parameters, which are key to safety issues in the context of transient or accidental reactor behavior. The Iterated Fission Probability (IFP) method allows the adjoint-weighted mean generation time and delayed neutron fraction to be computed within a Monte Carlo power iteration calculation. In this work we describe the specific features of the implementation of the IFP algorithm in the reference Monte Carlo code TRIPOLI-4 Ò developed at CEA. Several verification and validation tests are discussed, and the impact of nuclear data libraries, IFP cycle length and inter-cycle correlations are analyzed in detail.

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A new interactive homogenization procedure for reactor core or colorset calculations is proposed that requires iterative transport assembly and diffusion core calculations. At each iteration the transport solution of every assembly is used to produce homogenized cross sections for the core calculation. The converged solution gives assembly fine multigroup transport fluxes that preserve macrogroup assembly exchanges in the core. This homogenization avoids the periodic lattice - leakage model approximation and gives detailed assembly transport fluxes without need for an approximated flux reconstruction. In this paper we combined the benefit of a Domain Decomposition Method, that split the original transport problem in several multigroup fixed-source problems, with the effective solution of the Coarse-Mesh Finite-Differences operator that provides the whole-core eigenvalue and the neutron exchange between assemblies.

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