Test facility was designed in such a way that it enables to conduct experiments at any inclination of the flow channel. Test facility will be also used for demonstrational and educational purposes. Transparency of the channel walls and the use of WMSs offer students great possibilities to get better understanding on the physics behind the flow behavior. Test section is equipped with in-house manufactured WMSs (32 × 32 wires). The sensors are recording the flow at 5000 frames/s. In addition, special channel section was designed and constructed that enables High-Speed Camera and PIV measurements by minimizing the optical distortions.
To study the behavior of the PCCS configuration planned to be used in the ABWR II concept and to gain experimental data for the validation work of MELCOR severe accident code, a scaled down PCCS model was designed and constructed at Lappeenranta University of Technology in Finland in 2012–2013. The PCCS model is connected to the drywell and wetwell compartments of PPOOLEX, which is acting as a host facility. Steam needed in the experiments is produced with the nearby PACTEL facility. The PCCS model consists of five horizontal U-tube shaped heat exchange tubes installed inside a secondary side liquid pool. The pool is in atmospheric pressure and covered by a lid with an exit pipe out of the laboratory and with viewing windows for video cameras.
The PPOOLEX facility consists of a pressure vessel containing a wet well compartment (condensation pool), dry well compartment, inlet plenum and air/steam line piping. An intermediate floor separates the compartments from each other but a route for gas/steam flow from the dry well to the wet well is created by a vertical blowdown pipe attached underneath the floor.
PACTEL is a volumetrically scaled (1: 305) facility including a pressurizer, high and low pressure emergency core cooling systems, and accumulators. The reactor vessel is simulated with a U-tube construction including separate downcomer and core sections. The core itself is consists 144 full-length, electrically heated fuel rod simulators. Component heights and relative elevations correspond to those of the full scale reactor to match the natural circulation gravitational heads in the reference system. Three coolant loops with double capacity steam generators are used to model six loops of the reference power plant. The facility is still in operation for example as an auxiliary system for the separate effect test facilities. Until now, 239 experiments have been carried out with the facility.
The PWR PACTEL facility consists of a reactor pressure vessel model, two loops with vertical steam generators, a pressurizer, and emergency core cooling systems. The new loops and steam generators of EPR style construction enable the PWR and EPR related experimental research. The pressure vessel model in PWR PACTEL comprises a U-tube construction modeling the downcomer, lower plenum, core and upper plenum. The core rod bundle consists of 144 electrically heated fuel rod simulators arranged in three parallel channels. The core can be powered by a maximum of 1 MW electric power supply. The maximum core power corresponds roughly to the scaled residual heating power of the EPR reactor. The total height of the PWR PACTEL pressure vessel model corresponds to the pressure vessel height of EPR. The volume ratio between the pressure vessels in PWR PACTEL and EPR is about 1/405.
The LOBI Project originated from a reactor safety research and development contract between the European Commission and the former Bundesminister für Forschung und Technologie of the Federal Republic of Germany. On the basis of contingent and perceived safety requirements, BMFT LOBI Project decided in 1972 on the need of an experimental programme to be conducted in a integral system test facility to investigate thermal-hydraulic phenomenologies relevant to accident conditions in pressurized water reactors (PWRs) of German design. As result of a tender, the execution of this study was awarded to the Joint Research Centre of the European Commission.
Objectives:
- Investigation of Basic Phenomenologies governing the thermal-hydraulic response of an Integral System Test Facility for a range of PWR Operational and Accident Conditions.
- Provision of an Experimental Data Base for the Development and Improvement of Analytical Models and the Independent Assessment of Large System Codes used in LWR Safety Analysis.
Facility Configurations
The LOBI-MOD1 test facility configuration was designed to meet the relevant requirements originating from the objectives of the experimental investigations of Large and Intermediate Break Loss of Coolant Accidents (LOCA). A total of 28 tests were performed with this configuration during the period December 1979 to June 1982.
The LOBI-MOD2 test facility configuration, operating since April 1984, represents an upgraded version designed to meet also all relevant requierements related to the investigation of Small Break LOCAs and Special Transients. A total of 42 tests were performed in the period April 84 to June 1991.
Facility is dismanteled.
Test Matrix
In the PDF files attached below you can find more information on the Facility and a very useful Test Matrix with information of each experiment.
CERTA-TN
Also attached you may find information of CERTA - TN: European Thematic Network for the Consolidation of the Integral System Effect Experimental Databases for Reactor Thermal-Hydraulic Safety Analysis in which JRC participated with the LOBI Facility data and that was the origin to the development of the first version of the STRESA Database.
No description available.
ARIGS is one of the programs on the aerosol retention on the tubes surrounding the breach within the secondary side of the steam generator in the absence of water. Its development has been internationally framed within the EU-SGTR and the ARTIST programs. Experimental activities were focused on setting up a reliable database in which the influence of gas mass flow rate, breach configuration and particle nature in the aerosol retention are properly considered. Theoretical activities are aimed at developing a predictive tool (ARISG) capable of assessing source term attenuation in the scenario with reasonable accuracy. Given the major importance of jet aerodynamics, 3D CFD analyses are being conducted to assist both test interpretation and model development.
ARISG-I was a step forward in the modeling of the aerosol retention of the steam generator. According to this analysis the main areas where research is needed are: gas jet behavior across the tube bank; particle resuspension, erosion, and/or bouncing; and particle inertial impaction and turbulent deposition under foreseen conditions.
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
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 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.
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.
FOREVER program at KTH was concerned with the vessel integrity under the molten corium attack in the reactor lower plenum during a severe accident. Total 9 tests were performed in the FOREVER program, to simulate the behavior of the lower head of a reactor pressure vessel (RPV) under different conditions: French steel/American steel, with/without penetrations, with/without gap cooling.
The facility employs a 1/10th scaled lower head (hemispherical in shape and made of SA533B, American reactor steel) of 400 mm outer diameter and 15 mm wall thickness. A cylindrical shell of 15Mo3 German steel, of 400 mm height and thickness of 15 mm, was welded to hemispherical lower head to make a complete vessel.
Facility is in operation.
The SIMECO (SImulation of MElt COolability) test facility consists of a slice type vessel, which includes a semi-circular section and a vertical section, representing the lower head of the reactor vessel. The size of the facility is scaled to be 1/8 of prototypic PWR type reactors. Fig.1 shows a schematic of the facility and Fig.2 shows the main dimensions of the vessel test section. The diameter and height of the test section are 620 mm and 530 mm, respectively. The width of the test section is 90 mm. The front and back faces of the facility are insulated in order to decrease heat losses. The vessel’s wall, represented by a 23-mm thick brass plate, is cooled by a regulated water loop. On the top of the vessel a heat exchanger with regulated water loops is employed to measure the upward heat transfer. The sideways and downward heat fluxes are measured by employing array of thermocouples at several different angular positions. Practically isothermal boundary conditions are provided at pool boundaries. A cable-type heater 3 mm in diameter and 4 m in length is submerged in the pool and provides internal heating. A heat exchanger mounted on the exit of cooling water, is employed to maintain the cooling capacity of the water. The isothermal bath is designed to provide constant temperature. A circulation pump was mounted in order to establish necessary flow rate. One digital and one analog flowmeter were mounted to measure water flow through the wall of the slice, while one analog flowmeter is used to measure the flow in the upper heat exchanger.
Facility is in operation.
Reactor cavity flooding is a cornerstone of severe accident management strategy in Swedish type BWRs. In a hypothetical severe accident with core melting and reactor vessel melt-through, it is assumed that the melt ejected into a deep water pool will fragment, quench and form a porous debris bed coolable by natural circulation. If natural circulation cannot remove decay heat produced by the debris, then dryout, reheating and remelting of the debris bed is expected to occur. Attack of molten core materials on the reactor containment base-mat presents a threat to containment integrity. Amount of the heat which can be removed by natural circulation from the debris bed is contingent, among other factors, upon the properties of the bed as porous media. Debris agglomeration and especially formation of “cake” regions can significantly increase hydraulic resistance for the coolant flow and thus negatively affect coolability of the debris bed. If melt is not completely solidified prior to settlement on top of the debris bed, then agglomeration of the debris and even “cake” formation is quite possible.
The JRC-Ispra FARO plant is a large multipurpose test facility in which reactor severe accidents could be simulated by out-of-pile experiments. A quantity in the order of up to 200 kg of oxide fuel type melts (up to 3000 °C) could be produced in the FARO furnace, possibly mixed with metallic components, and delivered to a test section containing a water pool at an initial pressure up to 5.0 MPa. The reference scenario of the current test series is relevant to a postulated in-vessel core melt down accident when jets of molten corium penetrate into the lower plenum water pool, fragment and settle on the lower head.
Spreading: These tests are designed to investigate the impact on the core catcher of corium ejected after reactor pressure vessel failure during a core meltdown accident. The way melt spreads on the core catcher surface is important because of its effect on the long-term coolability of the melt. Two tests have been performed, one with a dry surface and one with 1 cm of water layer.
Objectives:
- investigation of basic phenomenologies relevant to the progression of severe accidents in water cooled reactors with particular emphasis on the interaction of molten fuel with coolant and/or structures under both in-vessel and ex-vessel postulated severe accident conditions.
- Provision of an Experimental Data Base for the Development and Improvement of Analytical Models and the Independent Assessment of Large System Codes used in LWR Safety Analysis.
Facility is dismanteled.
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 (Fuel-Coolant Interaction) phenomena relevant to severe accident situations in nuclear reactors.
KROTOS was transferred to CEA Cadarache at the end of the JRC-Ispra MFCI programme in 1999. It is at present part of the French institute research programme on severe accidents.
The JRC-Ispra FARO plant was a large multipurpose test facility in which reactor severe accidents could be simulated by out-of-pile experiments. A quantity in the order of up to 200 kg of oxide fuel type melts (up to 3000 °C) could be produced in the FARO furnace, possibly mixed with metallic components, and delivered to a test section containing a water pool at an initial pressure up to 5.0 MPa. The reference scenario of the current test series is relevant to a postulated in-vessel core melt down accident when jets of molten corium penetrate into the lower plenum water pool, fragment and settle on the lower head.
Quenching: Investigation of basic phenomenologies relevant to the fragmentation and quenching of molten material into the water coolant at different initial pressure and water subcooling. 12 Tests have been performed: 5 at 50 bar initial pressure, 1 at 20 bar and 6 tests at pressure lower than 5 bar. In the last test an external trigger was applied to the molten mixture.
Objectives:
- investigation of basic phenomenologies relevant to the progression of severe accidents in water cooled reactors with particular emphasis on the interaction of molten fuel with coolant and/or structures under both in-vessel and ex-vessel postulated severe accident conditions.
- Provision of an Experimental Data Base for the Development and Improvement of Analytical Models and the Independent Assessment of Large System Codes used in LWR Safety Analysis.
Facility is dismanteled.
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.
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.
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.
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.
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.
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.
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.
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 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.
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 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 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 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 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.