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.

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.

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 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.

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.