Facilities

The test section of the KROTOS facility consists of a stainless steel test section bolted to lugs welded on the inner side walls of a stainless steel pressure vessel. The cylindrical pressure vessel, inner diameter 0.4 m, height, 2.21 m, has a thick flat bottom and a flanged flat upper head and is designed to withstand a static pressure of 2.5 MPa at 493 K. The cylindrical test section, inner diameter 200 mm, outer diameter 240 mm, closed at the bottom by either a flat plate or with a gas trigger device, can contain water up to a height of about 1.27 m (about 40 litres).



The KROTOS main objective is to provide basic experimental information on FCI phenomena relevant to severe accident situations in nuclear reactors.

Facility is in operation at CEA. KROTOS was transferred to CEA Cadarache at the end of the JRC-Ispra MFCI programme in 1999.

For the JRC-Ispra KROTOS performed experiments see https://stresa.jrc.ec.europa.eu/facilities/krotos.

Determination of the vaporization rate according to the composition and the thermodynamic conditions of the corium (with FP simulants) was the aim of the COLIMA (COrium LIquid and MAterials) experiments. The facility provided representative conditions of the aerosols suspended inside the containment of PWRs under a severe accident. According to the scientific objectives of each experiment, different configurations of the facility can be used: corium/materials interaction (concrete, ceramics), release of aerosols from the corium (simulating physical-chemistry of oxidic and metallic fission products, without radioactive isotopes except uranium).

COLIMA consists of 1.5m3 tank, where the maximum internal pressure can reach 0.3MPa. The corium can be melted in a crucible by a thermite reaction or an induction coil that can maintain it hot in order to provide a steady state situation up to 3000◦C. The crucible, designed to contain few kilograms of corium, is surrounded by a thermal shield ring and can be placed at the bottom or at the middle of the tank. The walls of the vessel tank are thermally controlled at 150◦C. Portholes, dedicated to the instrumentation, are located at its top, half height and bottom.

VITI (‘‘VIscosity Temperature Installation’’) experimental assembly: (1) VITI chamber, (2) graphite crucible, (3) ZrCcoating, (4) studied mixture, (5) graphite susceptor, (6) thermal shield, (7) support for crucible, (8) support for thermal shield, (9) inductance coil, (10) pyrometer – measure of Tcrucible, (11) pyrometer – measure of Tmixture, (12) data acquisition

The experiments were dedicated to the selected coating interaction with water reactor corium and with sodium fast reactor corium compositions.
VITI facility has been developed to measure viscosity, density and surface tension on corium up to 2600 C by aerodynamic levitation. But it is also used as small crucibles heating for material interactions tests. Samples of less than 100 g can be studied in VITI.

In the hypothetical case of a nuclear reactor severe accident, the reactor core could melt and form a mixture, called corium, of highly refractory oxides (UO2, ZrO2) and metallic or oxidized steel, that could eventually flow out of the vessel and mix with the basemat decomposition products (generally oxides such as SiO2, Al2O3, CaO, Fe2O3, …).
The VULCANO experimental facility is operated to perform experiments with prototypic corium (corium of realistic chemical composition including depleted UO2). This is coupled with the use of specific high-temperature instrumentation requiring in situ cross calibration.
Due to the complex behavior of corium in the solidification range, an interdisciplinary approach has been used combining thermodynamics of multicomponent mixtures, rheological models of silicic semisolid materials, heat transfer at high temperatures, free-surface flow of a fluid with temperature-dependant properties.
Twelve high-temperature spreading tests have been performed and analyzed. The main experimental results are the good spreadability of corium–concrete mixtures having large solidification ranges even with viscous silicic melts, the change of microstructure due to cooling rates, the occurrence of a large thermal contact resistance at the corium–substrate interface, the presence of a steep viscosity gradient at the surface, the transient concrete ablation. Furthermore, the experiments showed the presence of the gaseous inclusions in the melt even without concrete substrate. This gas release is linked to the local oxygen content in the melt which is function of the nature of the atmosphere, of the phases (FeOx, UOy, …) and of the substrate. These tests with prototypic material have contributed to the validation of spreading models and codes which are used for the assessment of corium mastering concepts.
Facility is in operation.

In case of prolonged loss of cooling accident, the fuel rods of the core of a pressurized water reactor (PWR) will be damaged, and will collapse to form what is called a "debris bed", i.e. an agglomeration of fragments of zircaloy cladding and UO2 pellets (or UO2 and PuO2 pellets in the case of MOX fuel rods) which, if not rapidly cooled, will melt and become increasingly difficult to cool. This problem was identified through analysis of the Three Mile Island accident (TMI-2) which occurred in the United States in 1979.



One of the recommended actions to mitigate such accident sequences consists of reinjecting cooling water into the core, an action so-called "reflooding". Although essential for cooling the fuel assemblies, this action may nevertheless compromise the integrity of the reactor containment building. Indeed, reflooding a melting core at very high temperature may cause an explosive thermal reaction, so-called "steam explosion", between the cooling water and the molten corium. Such an explosion can generate projectiles which could damage the containment building. Furthermore, the water vapor resulting from the vaporization of the injected water will oxidize the metallic compounds of the core (zircaloy cladding, steel structures) and generate hydrogen with the potential to undergo a combustion inside the containment, as it was observed during the Fukushima accident.



The "Debris bed reflooding" experimental research program was launched in order to better understand and model these phenomena, the final objective being to determine the conditions under which cooling water can be injected so as to cool the core in an efficient manner with an acceptable risk for the containment. This additional knowledge will be subsequently used to clarify the choice of emergency operating procedures for severe accident conditions and to support the assessment of the relevance of EDF's Severe Accident Operating Guidelines.

Facility is in operation.

The tests conducted in the PRELUDE facility help to validate key technical options for PEARL:

• Induction heating to obtain heating sequences between 100-300 W/kg with homogeneous distribution in the different particle beds (slightly oxidised steel balls with 1, 2, 4 and 8 mm diameters), as well as to reach a temperature of 1,000°C at the hottest spot in the debris bed.
• Material of the test section ensuring the thermomechanical resistance of the tube containing the particles bed,
• Instrumentation to record the fi rst thermohydraulic measurements at atmospheric pressure when refl ooding the particle bed (about 25 kg) heated to of 400, 700 and 1,000°C.

This modular facility will remain operational to support the larger-scale PEARL facility (debris bed of about 500 kg) for complementary separate effects tests.

Facility is not operating, now called PEARL.