Analysis of Release Model Effect in the Transport of Fission Products Simulating the FPT3 Test Using MELCOR 2.1 and MELCOR 2.2

The severe accident integral codes such as Methods for Estimation of Leakages and Consequences of Releases (MELCOR) are complex tools used to simulate and analyse the progression of a severe accident from the onset of the accident up to the release from the containment. For this reason, these tools...

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Bibliographic Details
Published inSustainability Vol. 13; no. 14; p. 7964
Main Authors Flores y Flores, Alain, Ferretto, Danilo, Marková, Tereza, Mazzini, Guido
Format Journal Article
LanguageEnglish
Published Basel MDPI AG 16.07.2021
Subjects
Boron
Codes
Containment
Design
Experiments
Fission products
Nuclear accidents
Nuclear energy
Nuclear power plants
Nuclear reactors
Nuclear safety
Oxidation
Phenomenology
Pressurized water reactors
Radiation protection
Reactors
Simulation
Stainless steel
Sustainability
Czech Republic
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Summary:The severe accident integral codes such as Methods for Estimation of Leakages and Consequences of Releases (MELCOR) are complex tools used to simulate and analyse the progression of a severe accident from the onset of the accident up to the release from the containment. For this reason, these tools are developed in order to simulate different phenomena coupling models which can simulate simultaneously the ThermoHydraulic (TH), the physics and the chemistry. In order to evaluate the performance in the prediction of those complicated phenomena, several experimental facilities were built in Europe and all around the world. One of these facilities is the PHEBUS built by Institut de Radioprotection et de Sûrete Nucléaire (IRSN) in Cadarache. The facility reproduces the severe accident phenomena for a pressurized water reactor (PWR) on a volumetric scale of 1:5000. This paper aims to continue the assessment of the MELCOR code from version 2.1 up to version 2.2 underlying the difference in the fission product transport. The assessment of severe accident is an important step to the sustainability of the nuclear energy production in this period where the old nuclear power plants are more than the new reactors. The analyses presented in this paper focuses on models assessment with attention on the influence of B4C oxidation on the release and transport of fission products. Such phenomenon is a concern point in the nuclear industry, as was highlighted during the Fukushima Daiichi accident. Simulation of the source term is a key point to evaluate the severe accident hazard along with other safety aspects.
ISSN:2071-1050
2071-1050
DOI:10.3390/su13147964