Development and validation of ALEPH Monte Carlo burn-up code

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Abstract

The Monte-Carlo burn-up code ALEPH is being developed in SCKCEN since 2004. Belonging to the category of shells coupling Monte Carlo transport (MCNP or MCNPX) and "deterministic" depletion codes (ORIGEN-2.2), ALEPH possess some unique features that distinguish it from other codes. The most important feature is full data consistency between steady-state Monte Carlo and time-dependent depletion calculations. Recent improvements of ALEPH concern full implementation of general-purpose nuclear data libraries (JEFF-3.1.1, ENDF/B-VII, JENDL-3.3). The upgraded version of the code is capable to treat isomeric branching ratios, neutron induced fission product yields, spontaneous fission yields and energy release per fission recorded in ENDF-formatted data files. The alternative algorithm for time evolution of nuclide concentrations is added. A predictor-corrector mechanism and the calculation of nuclear heating are available as well. The validation of the code on REBUS experimental programme results has been performed. The upgraded version of ALEPH has shown better agreement with measured data than other codes, including previous version of ALEPH.

Details

Original languageEnglish
Title of host publicationNuclear measurements, Evaluations and Applications - NEMEA-6
Place of PublicationParis, France
Pages161-170
Publication statusPublished - 11 Aug 2011
EventNuclear measurements, Evaluations and Applications - NEMEA-6 - Krakow, Poland
Duration: 25 Oct 201028 Oct 2010

Conference

ConferenceNuclear measurements, Evaluations and Applications - NEMEA-6
CountryPoland
CityKrakow
Period2010-10-252010-10-28

Keywords

  • burn-up, nuclear data, differential equations

ID: 162327