ARTEMIS Program : Investigation of MCCI by Means of Simulating Material Experiments
- Commissariat a l'Energie Atomique (CEA), DTN/SE2T/LPTM, 17 rue des Martyrs, 38054 Grenoble (France)
In the frame work of R and D on Severe Accidents in PWR plants, an estimation by codes of time of basemat melt-through by corium is required. For this, the heat flux distribution along the cavity wall must be properly modelled. Hence the knowledge of the heat transfer coefficient as well as the temperature at the interface between the melt and the solid become key issues. The phase diagram of the melt and composition governs the interface temperature which controls, at least partly, the thickness of the corium crust formed on the molten concrete. Crust behaviour (time evolution of thickness, mechanical interaction with gas) implies a release mode of molten concrete in Corium which in turn alters the melt composition. Clearly, the molten corium-concrete interaction (MCCI) phenomenon is the result of a strong coupling between physico-chemistry and thermal-hydraulics. The main goal of the first test series of the ARTEMIS program, essentially sponsored by the Institut de Radioprotection et de Surete Nucleaire (IRSN) is to make a link between the interface temperature and the physico-chemistry of the melt (phase diagram) through tests conducted with simulating materials and to provide an insight on the existence, the behaviour and the composition of the crust. This test series considers 1D MCCI using a non eutectic LiCl-BaCl{sub 2} mixture poured at 1000 deg C in a cylindrical test section (internal diameter 0.3 m) to interact with the 0.35 m deep basemat made of the same salt mixture at the eutectic composition. This 'concrete' was especially manufactured with sintered granulates to allow gas flow from the bottom (argon), then simulating gas released by concrete in the reactor case. Constant power is applied in the pool with a helical coil and 1D MCCI is ensured by counterbalancing heat losses by controlled heating at the lateral walls and at the top of the test section. Concrete ablation is followed from the output of 45 0.5 mm diameter thermocouples. An instrumented rod periodically investigates the temperature and the position of the interface between corium and solid. At last, measurements of the corium composition during the test as well as post-mortem analysis are implemented. In this paper, similarity criteria for phase diagram and thermal-hydraulics to select adequate simulating material are first described. Then the most salient results obtained during the 6 tests performed are given and according to these results, a classification in 3 categories is proposed. It is concluded that for tests submitted to conditions prevailing in the reactor the experimental results agree satisfactorily with the essential features of the so-called 'phase segregation model' envisaged in the TOLBIAC-ICB MCCI code. (author)
- Research Organization:
- American Nuclear Society, 555 North Kensington Avenue, La Grange Park, IL 60526 (United States)
- OSTI ID:
- 21016379
- Resource Relation:
- Conference: 2006 International congress on advances in nuclear power plants - ICAPP'06, Reno - Nevada (United States), 4-8 Jun 2006; Other Information: Country of input: France; 8 refs; Related Information: In: Proceedings of the 2006 international congress on advances in nuclear power plants - ICAPP'06, 2734 pages.
- Country of Publication:
- United States
- Language:
- English
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