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.

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.

No description available.

Steam generator reliability and performance are serious concerns in the operation of pressurized water reactors. The aim of the SGTR project was to provide a database of fission product retention in steam generator tube rupture sequences and models, which could be applied to estimate the effectiveness of different accident management strategies in these kind of accidents.
The SGTR project made an important step forward to resolve uncertainties of physical models, especially in the aerosol deposition and mechanical resuspension in turbulent flows. There was one sampling at the injection line for the Optical Particle Counter (OPC) aimed at determining the aerosol size distribution and quantifying the mass concentration at the inlet. Within the vessel atmosphere eight samplings were taken to six filters and two cascade impactors, from which the mass concentration exiting the tube mini-bundle was estimated.
The test mini-bundle is a scaled mock-up of the first stage of the steam generator tube bundle. It consists of a squared arrangement housing inside a total of 117 tubes plus four supporting rods placed in the corners. The mini-bundle allows two possible locations of the broken tube. One place is just at the centre of the structure and the other place is three tubes away from the centre.

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 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.