[en] The Phebus FPT3 experimental test is the last in-pile integral test of the Phebus F.P. program which is dedicated to investigate fuel rod degradation and behaviour of Fission Products (FPs) released via the primary coolant circuit into the containment building. The purpose is to provide new experimental results to validate the different codes used in water reactor safety analysis. Unlike the previous tests, FPT3 used boron carbide as absorber material inserted in the pre-irradiated fuel bundle. The analysis of the FPT3 data presented in this paper highlights some major events occurring during the fuel bundle degradation like fuel clad and absorber rod failures, material relocation and hydrogen and carbonaceous species production kinetics. Moreover, the final bundle state was intensely characterised by a series of post test non destructive examination in order to give additional information on the material movements in the fuel bundle. Besides, g-measurements performed on-line and off-line on the sequential samplings from the containment vessel provide data for the description of the gaseous iodine behaviour throughout the test. Finally, hydrogen recombiner coupons (which are placed in the reactor containment buildings to prevent the local hydrogen accumulation) were exposed inside the containment soon after the degradation phase to evaluate their efficiency in conditions representative of a reactor severe accident. (orig.)

[en] The purpose of this paper is to provide a preliminary overview of the phenomena observed during the experimental phase of the PHEBUS Fission Product Test FPT3. This experiment was the last in the series of 5 in-pile integral experiments performed by IRSN in the PHEBUS facility operated by the CEA on the site of Cadarache. Unlike the previous tests, FPT3 used boron carbide as absorber material instead of silver-indium-cadmium, so varying an important parameter impacting physico-chemical phenomena. FPT3 test course was in agreement with the pre-defined test protocol, including a 8,5-day irradiation phase, a fuel bundle degradation phase which lasted less than 5 hours and a 4-day long-term phase that consisted of an aerosol stage dedicated to the analysis of aerosol deposition mechanisms inside the containment vessel and a chemistry stage devoted to the analysis of the iodine chemistry. During the experiment, both the on-line instrumentation and the periodic samplings worked quite well. The fuel degradation progress could be analysed through both temperatures inside the bundle and gaseous concentration measurements performed in the circuit and inside the containment vessel. Some major events, like fuel clad and absorber rod failures or material relocations, were clearly correlated to both bundle and circuit instrumentation signals. The post test non destructive examinations of the fuel bundle (X-radiography, X- and γ-tomographies and γ-scanning) allowed to compare FPT2 and FPT3 bundle final degradation states. On-line γ-detector measurements coupled with numerous post test gamma-counted sequential samplings help for the characterization of the iodine behaviour inside the containment vessel during the degradation and the long term phases. The whole set of measurements appears self-consistent and provides new data for the iodine solubility inside the sump, the iodine gaseous fraction and the organic versus molecular iodine distribution inside the containment atmosphere. (authors)

[en] Fuel-coolant interaction (FCI) is an important issue for the assessment of severe accident safety for both sodium-cooled fast reactors (SFRs) and pressurized water reactors (PWRs). For the ASTRID SFR demonstrator, FCI is a key phenomenon affecting the relocation of molten fuel in engineered discharge tubes between the core region and the core catcher plenum. FCI controls jet fragmentation and debris bed formation and raises the issue of potentially energetic vapor explosions in the ASTRID lower head. In this frame, experimental data will be necessary to validate SCONE, the fuel-sodium interaction code under development at CEA. For PWRs, one of the configurations of interest lies within the residual case where in-vessel retention would fail. In this case, it is expected that a light metallic layer would be the first to interact with water, before a heavier oxide melt discharge. Here, steam explosion and debris bed formation are the two major points of interest. Based on the experimental expertise gained from the KROTOS facility and its X-ray radioscopic imaging system, new test facilities have been designed to carry out prototypic (depleted uranium-containing) corium interactions with either sodium or water in PLINIUS2, the CEA future large-mass experimental platform dealing with masses above 100 kg. Some test sections have been specially designed to ensure proper visualization of the fuel, liquid coolant, and vapor phases by an improved X-Ray imaging system. This paper presents the future PLINIUS 2 platform as well as the experimental programs foreseen to study both water-corium and sodium-corium interactions. (authors)