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

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 facility models the reactor pressure vessel (RPV), part of the reactor cooling system (RCS), the cavity, and the pump and steam generator rooms. The length scale is 1:18, compared to a large European reactor.

The pressure vessel consists of a steel pipe with a model of the RPV (outer diameter 298.5 mm) at its lower end. It has a total volume of 0.0879 m³, that models the volume of the pressure vessel and the volume of part of the RCS. The lower head of the RPV can hold 3.4 10-3 m³ of liquid, which corresponds to 20 m³ or 160 t of corium.



The cavity, a Plexiglas cylinder with an inner diameter of 342 mm, is attached to the vessel support structure. Generally, the flow path out of the reactor pit is through the annular gap between reactor vessel and cavity wall and along the main cooling lines into the pump and steam generator compartments. There is the option of a flow path through holes straight up into the containment, that is modeled by an extra cylindrical compartment.



The compartments, eight boxes which model in volume the steam generator and pump rooms (0.3 m³ and 0.131 m³ each respectively), are connected to the nozzles, and are placed on the vessel support structure around the RPV. They are covered by filters on their tops for the extraction of fog and drops. Two boxes have one Plexiglas wall each, to permit optical access for flow visualization.



The following failure modes of the lower head were studied: central holes and three types of lateral breaches: lateral holes, a horizontal slot, and complete ripping and tilting of the lower head. The horizontal slot models a partial rip in the lower head, as it might occur with a sidepeaked heat flux distribution. The flow cross section is equivalent to a 25 mm hole. The fluids employed were water or a bismuth alloy (similar to Wood’s metal) instead of corium, and nitrogen or helium instead of steam. Most experiments were performed for the combination water/nitrogen, with 3.410-3 m3 of water. With central holes four hole sizes, 15, 25, 50 and 100 mm diameter (scaled 0.27 m – 1.80 m), were investigated, each at three initial pressures, 0.35, 0.6 and 1.1 MPa. Nitrogen/metal tests were performed with the 25 and the 50-mm-hole at 0.6 and 1.1 MPa with 3.310-3 m3 of metal. Some tests (25 and 50 mm hole size) were performed with 1.810-3 m3 of water and were repeated several times. The reproducibility was very good.

Facility is in operation.

The main components of the facility are scaled about 1:18 linearly to a large European reactor: the containment pressure vessel (volume 14 m³), the RPV-RCS pressure vessel (0.08 m³), the cavity, the subcompartment, and the steam accumulator (0.08 m³). The subcompartment is an annular space around the cavity with a volume of 1.74 m³. The flow path from the cavity into the subcompartment is along the eight stubs modeling the main cooling lines (total flow cross section is 0.0308 m²). The connection from the subcompartment to the containment is by four openings with a diameter of 130 mm in its top plate. The RCS-RPV pressure vessel models the volumes of both the reactor cooling system (RCS) and the reactor pressure vessel (RPV). The RPV model, that serves as crucible for the generation of the melt, is bolted to a plate carrying the RCS-RPV pressure vessel. The hole at the bottom of the melt generator is formed by a graphite annulus. It is closed with a brass plate. The reactor pit is made of concrete and is installed inside a strong steel cylinder. Besides the flow path along the main cooling lines there is the option of a flow out of the cavity straight up into the containment through eight openings with a total cross section of 0.052 m². Depending on the reactor design, that is to be investigated, this cross section is variable.



In case of the modeling of a prototypical scenario, the containment vessel is heated by filling with steam additional to the atmospheric air until the pressure reaches 0.2 MPa. The average gas temperature and the wall temperature inside the vessel is 373 K (100°C) at the end of the heat-up. A metered amount of hydrogen gas (3 mol-%) is added to the vessel at the end of heat-up while fans are running inside the vessel. A gas sample is taken just before the start of the experiment.



The pressure vessel modeling the RPV and RCS volume is electrically heated to the saturation temperature of the planned burst pressure, e.g. to 453 K (180°C at 1.0 MPa). It contains nitrogen at that temperature at 0.1 MPa. The steam accumulator is heated electrically to the saturation temperature of twice the planned burst pressure, e.g. 486 K (213°C at 2.0 MPa). The accumulator is filled with a measured amount of water by a high pressure metering pump to reach that pressure. The RCS pressure vessel and the accumulator are connected by a 25 mm diameter pipe with an electro-pneumatically actuated valve.


The model of the RPV is filled with aluminum-ironoxide thermite. The experiment is started by igniting the thermite electro-chemically at the upper surface of the compacted thermite powder. When a pressure increase in the RPV-RCS pressure vessel verifies that the thermite reaction has started, the valve in the line connected to the accumulator is opened and steam enters the pressure vessel. When the pressure has reached a preset value the valve is automatically closed again. About 5 to 8 seconds after ignition the brass plug at the bottom of the RPV vessel is melted by the 2400 K hot iron-alumina mixture. That initiates the melt ejection. The melt is driven out of the breach by the steam and is dispersed into the cavity and the containment.

The KAJET erosion test facility is shown in Fig. 10-1. Total melt masses of up to 300 kg can be provided by various types of melt generators. Driving pressures of up to 2.5 MPa can be established. Melt release occurs downward into a vessel which is 1100 mm in diameter and 1900 mm in height and has at its bottom layers of gravel and sand. The pressure inside the vessel can be raised up to 0.3 MPa. The examined samples consisted of siliceous concrete and borosilicate glass concrete. The schematic (Fig. 10-2) helps to explain how the test was conducted. The time scale begins with the start of ejection. The first melt component to be ejected on sample no 1 was iron. Shortly before the end of iron release, the plate carrier was turned by 90° within one second. During that time, the melt changed to oxide as the component to be ejected on sample no 2.

The test facility consists of a slender cylinder with an effective inner diameter of 0.66 m bearing plane glass windows in the front and back sides. The upper part is occupied by the melt generator which leaves a 0.14 m wide annular space for the steam flow. The facility is entirely closed except for four large venting pipes (4 m long, each with a cross section of 90 cm2) which were also closed in two tests. The space below the melt generator is about 2.2 m high but for the processes to be studied, the actual height of the test water pool was determined by a concave debris catcher that could be mounted at different heights. The test rig was placed inside a large (220 m3) pressure vessel providing a safety barrier and the possibility to perform tests at elevated ambient pressure.

The QUEOS facility consists of the test vessel, the furnace and the valve system separating the two. The spheres are heated in an electric radiation furnace in an argon atmosphere. The spheres are discharged into the water with a drop height of 130 cm. The diameter of the sphere stream is 100 mm or 180 mm after the discharge from the middle valve and the spheres fall freely without touching any walls.

Facility is in operation.

The ECO facility, housed inside the large FAUNA steel vessel, was designed for investigating energy conversion ratios up to 20 %, related to 10 kg of melt. In principle, it consists of a piston/ cylinder system having 4 m in the total height. The melt generator providing the melt (≈ 2600 K) is part of the cylinder. The water pool has 0.6 m in diameter, and up to 1.3 m in depth. The steam explosion is triggered by two explosion capsules located in the centre and at the edge of the bottom of the pool. The test vessel (”piston”), under the pressure forces developing due to the steam explosion, moves downwards against the resistance of the underlying crushing material.

An important accident management measure for controlling severe accident transients in light water reactors (LWRs) is the injection of water to cool the degrading core. Flooding of the overheated core, which causes quenching of the fuel rods, is considered a worst-case scenario regarding hydrogen generation rates which should not exceed safety-relevant critical values. Before the water succeeds in cooling the uncovered core, there can be an enhanced oxidation of the Zircaloy cladding that in turn causes a sharp increase in temperature, hydrogen production, and fission product release. The complex physico-chemical processes during quenching and their mutual influence is not yet sufficiently known. In most of the code systems describing severe fuel damage the quench phenomena are only modeled in a simplified empirical manner.

Facility is in operation.

The LIVE facility at the Forschungszentrum Karlsruhe is designed to study the late phase of core degradation, onset of melting and the formation and stability of melt pools in the RPV. Additionally, the regaining of cooling and melt stabilisation in the RPV by flooding the outer RPV or by internal water supply have been investigated.
The experimental programme consists of three different phases. In the first phase (LIVE1) the investigations concentrate on the behavior of a molten pool, which is poured into the lower head of the RPV taking into account possible 3-d effects. The objective was to determine the time dependent local heat flux distribution to the lower head, and the development of crusts, depending on internal melt heating and external cooling modes. The gap formation between the RPV wall and the melt crust as well as the role of phase segregation of a non-eutectic, binary melt on the solidification behavior has been investigated. In the second phase (LIVE2) the experiments were extended to allow multiple melt pours and the presence of water in the lower head. The third phase (LIVE3) dealt with processes during in-core melt pool formation, the stability of the melt pools in the core region during different cooling modes and relocation processes after crust failure.
The experiments have been carried out with different simulant materials. The first melt was a binary mixture of NaNO3 and KNO3 with temperatures up to about 350 °C. In an advanced stage, the second melt was a binary mixture of V2O5 with CuO, MgO or ZnO with temperatures up to 900 °C.
Experiments in the LIVE facility were part of the LACOMERA Project of the EU 5th Framework Programme. Produced experimental database has been used to validate and improve computer models, which had being developed in the area of molten pool formation and cooling in the lower head.
Facility is in operation.