Facilities

The PASTELS project will demonstrate how innovative passive safety systems can support modernisation and optimisation of the European nuclear industry by providing relevant new safety options. The overall objective of the project is to improve the ability of European nuclear actors to design and deliver innovative passive safety systems - which are particularly promising as they do not rely on power supply or human intervention - and simulate their behaviour to support the safety demonstration.

PASTELS will make significant progress in the study of two specific passive systems, the Containment Wall Condenser (CWC) and the Safety Condenser (SACO) by:

• Building on and leveraging existing available computational codes to simulate the relevant thermal-hydraulic phenomena,
• Developing a robust, validated, multi-scale simulation methodology of passive systems,
• Performing new experimental studies to obtain the relevant validation data.

The project will deliver extensive methodology guidelines as well as a roadmap to achieving the licensing and implementation of these innovative passive system technologies in future European Nuclear Power Plants (NPPs).

EU FASTNET PROJECT CALCULATIONS

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.

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.

The test section was located downstream of the mixing vessel; it consisted of four steel pipes connected in series and/or parallel. The first pipe between the mixing vessel exit and the test pipe inlet (total length ≈4 m) was thermally insulated in order to reduce thermophoretic deposition and heat losses as well as to avoid steam condensation. The 63-mm inner diameter test pipe was 5 m long and was surrounded by an oven to keep the pipe wall temperature at the required levels during the deposition and resuspension phases. In the deposition phase, the carrier gas and aerosols pass through the mixing vessel a first straight pipe into the test section and then straight to the wash and filtering system. In the resuspension phase, the clean gas was injected through the resuspension line directly into the test section and the resuspended aerosols were collected in the main filter before the gas goes through the wash and filtering system.

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.

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 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 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.