Facilities

The activities covered the following three areas:

• Thermal hydraulic calculations described the cooling conditions possibly established during the incident.
• Simulation of fuel behaviour described the oxidation and degradation mechanisms of fuel assemblies.
• The release of fission products from the failed fuel rods was estimated and compared to available measured data.

The produced numerical results improved the understanding of the causes and mechanisms of fuel failures during the Paks-2 incident and provided new information on the behaviour of nuclear fuel under accident conditions.

The RUSET experimental programme was launched in 2002 at the Hungarian Academy of Sciences KFKI Atomic Energy Research Institute (AEKI). The aim of the program was to get data for assessment of ruthenium release at severe accident with air ingress. More than forty small scale tests have been performed with mixed powder components of inactive materials and with short fuel rods. The influence of temperature, air flow rate and the presence of other fission products on the gaseous Ru release and the retention role of fuel pellets and cladding have been investigated. The test series indicated that if an air ingress type severe accident occurs most of the initial Ru mass can be released from the reactor core to the containment or environment. Some part of the released gaseous Ru undergoes precipitation and deposits on the cold surfaces, another part is released in gaseous form. The deposited Ru oxides can serve as a secondary source for further gaseous Ru release.

Since the beginning of the 1990s’, several experimental series have been performed at the Hungarian Academy of Sciences KFKI Atomic Energy Research Institute (AEKI) with E110 and Zircaloy claddings. The aims of these experiments were to study and to compare the mechanical properties of the cladding materials and to investigate the effect of oxidation and hydrogen uptake on the mechanical performance of the claddings under accident conditions. The objectives have been achieved through separate effect tests with well defined conditions.



The main results of the tests have been collected into the Experimental Database of E110 Claddings under Accident Conditions. The database involves the data of several experimental series in the following main directories:

• Cladding ballooning tests (more than 170 tube tests).
• Tensile tests with tube and sheet specimens (more than 100 samples).
• Oxidation tests between 500-1200 °C.
• Ring compression tests (more than 100 samples).

Most of the tests were carried out with E110 cladding, but several experimental points were produced with Zircaloy-4 cladding as well for comparison purposes. The database includes not only the on-line measured data, but the results of post-test examinations (visual observations, metallographic analysis, SEM analysis). Experimental technical reports and some selected papers describing the tests are also available in the database.

The high temperature interaction of reactor core materials during a severe accident leads to the oxidation and melting of metal components. The interactions of molten Zr cladding with uranium-dioxide and zirconium-dioxide are important factors in the determination of fuel failure conditions and play a role in the loss of fuel rod like geometry and in the formation of debris bed and molten pool in the core.



The dissolution of uranium-dioxide and zirconium-dioxide by molten Zircaloy were investigated in earlier separate effect tests. UO2 dissolution experiments were carried out with UO2 crucible and Zr charge. The simultaneous UO2-ZrO2 molten Zircaloy dissolution was investigated in UO2 crucibles with Zircaloy charge and ZrO2 central rod [1]. The analysis of different experiments showed some discrepancy between the results, which was connected with different crucible sizes, UO2/Zr mass ratios and the melt surface to volume ratios [2]. The comparison of simultaneous dissolution tests of UO2/ZrO2 and separate UO2 and ZrO2 showed faster dissolution and larger extent of dissolution in the case of simultaneous tests. These observations emphasized the importance of prototypic conditions on the dissolution process. For these reasons experiments with short fuel rod segments were carried out in the KFKI Atomic Energy Research Institute, Budapest. The development of dissolution models needs more prototypical experiments, this task was addressed in the COLOSS project of the 5th FWP. The analytical support of the AEKI short fuel rod dissolution tests was provided by COLOSS partners [3],[4]. The results of the experimental series are expected to make possible the further model development and code validation. 2. OBJECTIVES

VERDON programme has been launched by the CEA as a follow-up of VERCORS programme. It addresses the consequences of a degradation of fuel elements in contact with air following penetration of the vessel after the meltdown of part of the reactor core or the dewatering of a spent fuel storage pit, especially the release and chemical behaviour of ruthenium (tests of release of fission products have been held under EPICUR programme as well).

The data base on Ru release under air ingress conditions from irradiated PWR fuel rods was still scarce, as in the VERCORS programme, few tests have been performed in very oxidising conditions and more particularly under air ingress with significant amount of air. In this context, VERDON programme included specific air ingress test on a genuine irradiated UO2 fuel sample in its original cladding. As in VERCORS programme, the sample has been previously reirradiated at low power in a MTR reactor, in order to rebuild the inventory of short halflife fission products (including 103Ru). This test has been conducted in a new dedicated hot cell. The aim was not only to measure the release of fission products, but also to study their deposit on thermal gradient tubes and their potential revolatilisation induced by air injection. Compared to VERCORS, VERDON included by more detailed examinations of the fuel sample before and after the tests, using microanalytical techniques, such as SEM, EPMA and SIMS in order to determine the location of the fission products within the various phases as well as the corresponding compounds if possible. This gave better understanding of the mechanisms, which promote fission products release in such situations, as well as supported the associated modelling. VERDON programme is a part of the International Source Term Programme, which is composed of separate effect tests aiming at reducing uncertainties in severe accident analyses.
Facility is in operation.

Complex measurements were performed at the integral high temperature test facility CODEX (COre Degradation Experiment) between 1995-2002 with electrically heated UO2 fuel rod bundles. The test matrix included the first VVER-440 type integral severe accident experiment. The results of a quench test with pre-oxidised bundle indicated the protective role of the external oxide scale. Unique experiments were performed with PWR bundles under air ingress conditions. The last test of the current series helped to resolve the methane production issue during the oxidaton of a boron-carbide control rod in a severe accident. Some experiments were related to the preparation of PHEBUS tests, and some others were performed parallel with similar QUENCH tests. The experimental results contributed to the general understanding of severe accident progression in the loss of rod-like geometry phase and the test data have been used and are available for model development and code validation purposes.

The CODEX out-of-pile integral test facility was built and put into operation in 1995 at the KFKI Atomic Energy Research Institute, Hungary in order to investigate some specific aspects of core degradation and to extend the experimental database for code valiadation and development. Some of the experiments were VVER specific, while others were of general interest for any light water reactor. The comparison of CODEX exepriments with CORA and QUENCH tests can help to sift out the effects related to the specific features or scaling of the facilities. Some new techniques (e.g. aerosol measurements) applied in the test facility provide additional information on the high temperature behaviour of core materials. For the investigation of the aerosol release a cascade impactor system is connected to the upper plenum of the cooler and two pipelines allowes the continuous measurement of aerosols by means of laser particle counters. The gas concentration in the off-gas system is measured using a quadropole mass spectrometer. The instrumentation of the facility consists of the measurements of the operational parameters as heating power, flowrates, temperatures, levels and pressures. Thermocouples are placed in several positions in the heat insulation material, on the heat shield, on the external surface of the shroud, on the fuel rods and inside of the central (unheated) rod. Two pyrometers and a video camera are located at three windows in the upper part of the bundle. After the experiments the post-test examination of the bundle and aerosol samples is carried out with several techniques, including metallography, SEM, microprobe analysis, X-ray radiography and mass spectrometry.
Facility is not in operation.

The experiment objective was to study the physical phenomena that affect hydrogen distribution in the reactor containment such as: steam wall condensation, heat mass and momentum exchanges with the sump or with the containment spray systems. These different phenomena have been studied during specific test phases.
TOSQAN facility is highly instrumented both in terms of measurement density and diversity. Most of instrumentation is based on innovative optical diagnostics, which allows to measure accurately and non-intrusively the multiphase flow composed of various gases (air, steam, and helium used as a surrogate of hydrogen), water droplets, and aerosols simulating the fission products.
Facility is in operation.

The influence of containment sprays on atmosphere behaviour is being investigated both experimentally and theoretically. Experiments are being performed on the TOSQAN and MISTRA experimental facilities. The main objective of the CEA's MISTRA programme was to study condensation on the walls and the water droplets (from spraying) in a geometry larger than that of TOSQAN and with the possibility of compartments.
The experiments, carried out at MISTRA within SARNET, followed the same basic pattern. First, a well-defined (in terms of pressure, temperature and atmosphere composition) initial state was obtained, with a quiescent atmosphere. Then, sprays were activated with all boundary conditions remaining constant. The tests lasted typically less than two hours.
Facility is in operation.

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.