The european JASMIN project for the development of a new safety simulation code, ASTEC-Na, for na-cooled fast neutron reactors
| Autor: | Girault, N., Dorsselaere, J. P., Jacq, F., Brillant, G., Kissane, M., Bandini, G., Buck, M., Champigny, J., Hering, W., Perez-Martin, S., Luis E. HERRANZ, Raison, P., Reinke, N., Tucek, K., Verwaerde, D. |
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| Zdroj: | Scopus-Elsevier |
| Popis: | The 4-year JASMIN collaborative project, involving 9 organizations, was launched by IRSN end of 2011 within the 7th European R&D Framework Programme on the enhancement of Na-cooled Fast Neutron Reactors (SFR) safety for a higher resistance to severe accidents. The project aims at developing a new European simulation code, ASTEC-Na, with a modern architecture, sufficiently flexible to account for innovative reactor designs and eventually new types of fuel and claddings and accounting for results of recent research outcomes on water-cooled reactors. The code will be based on the ASTEC European code system, developed by IRSN and GRS for severe accidents in water-cooled reactors, and it will integrate and capitalize the state-of-the-art knowledge on SFR accidents through the improvement of existing physical models or the development of new ones. The code objectives will be to predict throughout the primary phase of the accidental sequence the cladding and fuel behaviour as well as the behaviour of the released fission products both in the primary circuit and in the containment vessel, including the extreme thermal-hydraulic conditions prevailing in case of Na fire. The main involved phenomena that will be investigated during the project include fuel element heat-up and failure, fuel-coolant-interaction, fuel dispersion or compaction, neutronics of the degraded core, fission product retention by sodium pool scrubbing, sodium aerosol depletion and physico-chemical transformations in the containment vessel. The first 18 month period of the project is mainly dedicated to build model specifications as well as code validation matrices related to fuel pin degradation in transient events and in-containment phenomena. The developed models will be validated as far as possible on existing in-pile (mainly the past CABRI experiments) and out-of-pile experimental data. JRC.F.5-Nuclear Reactor Safety Assessment |
| Databáze: | OpenAIRE |
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