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.
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.
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 catalyst sheets (stainless steel coated with washcoat/platinum catalyst material) are arranged in parallel forming vertical rectangular flow channels. Such a set-up represents a box-type recombiner section of AREVA design. Inside of the configuration the distribution of the catalyst temperatures and the gas com- position in vertical flow direction are measured. The correlation of the hydrogen conversion and catalyst temperatures with the experimental parameters serve basically to clarify the interactions of reaction kinetics, heat and mass transfer, and the flow conditions inside the recombiner.
Facility is 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.
In order to comply with experimental device design requirements, different devices were developped, tested and set up on CARAIDAS:
- experimental enclosure in which representative thermodynamic conditions could be achieved,
- the monosized drops generator, the drops diameter measurements and the drops collector,
- the cesium iodide aerosols generator, concentration and size distribution measurements.
Facility is not operating.
TUBA (thermo-phoresis and diffusiophoresis) programme included experiments represented conditions expected in the SG tubes. Laminar flow was used in TUBA tests (TUBA-T: thermophoresis and TUBA-D: diffusiophoresis & thermo-diffusiophoresis).
From the comparison of experimental results and SOPHAEROS calculations it was concluded that there was sufficient agreement for most of the cases studied.
Facility is dismanteled.
RECI is a 2½ year experimental programme that was brought to completion as of October 2004. The aim of the RECI (RECombiner & Iodine) program was to quantify the iodide → iodine conversion in realistic conditions of recombiner operation, albeit under the following constraints: the experiments were to be performed with non-radioactive substances, and without hydrogen. The comprehensive tests grid allowed to investigate into the decomposition of cesium and cadmium iodides under thermal-hydraulics conditions that mimics the recombiner operation, despite the technical limitations of the RECI test bench.
The aerosol generator selected limits the RECI programme to the study of water soluble substances, namely cesium and cadmium iodides: silver iodide is insoluble in water, and indium monoiodide is hydrolysed. However, the experimental results can be interpolated with reasonable confidence, since CsI and CdI2 are the two end-terms in the stability range of the relevant iodides. The instability of metal iodides, in a wet and oxidizing atmosphere, already demonstrated in chemistry laboratories, has been confirmed in more relevant physico-chemical conditions. The high conversion yields obtained do not come as a surprise since the RECI experiments provide a close analogy to the processes known as spray drying and spray (reactive-, or oxidizing-) pyrolysis, widely used in the laboratory and in the manufacturing industry. Both processes capitalize upon the high surface/volume ratio of aerosol particles to master comparatively slow chemical reactions and to produce nano-particles, the precursor material being often a finely powdered metal halide.
The experimental test bench consists of 4 units.
- Aerosol generation: an ultrasonic aerosol generator atomises the aqueous solution of a water soluble iodide. Monodispersed droplets are then dried, yielding iodide particles, the size of which is determined by the concentration of the solution. The input power of the ultrasonic acoustic transducer sets the aerosol concentration.
- Recombiner surrogate: a clear fused quartz or alumina tubing, which can accommodate a catalyst foil, is heated in a vertical tube furnace.
- Aerosol characterization: particles concentration & size distribution measurements. • Gaseous iodine analysis: 3 independently calibrated methods are implemented in the flue gas.
Facility is dismanteled.