Facilities

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.

Complex measurements were performed at the integral high temperature test facility CODEX (COre Degradation Experiment) between 1995-2002 with electrically heated UO2 fuel rod bundles. The test matrix included the first VVER-440 type integral severe accident experiment. The results of a quench test with pre-oxidised bundle indicated the protective role of the external oxide scale. Unique experiments were performed with PWR bundles under air ingress conditions. The last test of the current series helped to resolve the methane production issue during the oxidaton of a boron-carbide control rod in a severe accident. Some experiments were related to the preparation of PHEBUS tests, and some others were performed parallel with similar QUENCH tests. The experimental results contributed to the general understanding of severe accident progression in the loss of rod-like geometry phase and the test data have been used and are available for model development and code validation purposes.

The CODEX out-of-pile integral test facility was built and put into operation in 1995 at the KFKI Atomic Energy Research Institute, Hungary in order to investigate some specific aspects of core degradation and to extend the experimental database for code valiadation and development. Some of the experiments were VVER specific, while others were of general interest for any light water reactor. The comparison of CODEX exepriments with CORA and QUENCH tests can help to sift out the effects related to the specific features or scaling of the facilities. Some new techniques (e.g. aerosol measurements) applied in the test facility provide additional information on the high temperature behaviour of core materials. For the investigation of the aerosol release a cascade impactor system is connected to the upper plenum of the cooler and two pipelines allowes the continuous measurement of aerosols by means of laser particle counters. The gas concentration in the off-gas system is measured using a quadropole mass spectrometer. The instrumentation of the facility consists of the measurements of the operational parameters as heating power, flowrates, temperatures, levels and pressures. Thermocouples are placed in several positions in the heat insulation material, on the heat shield, on the external surface of the shroud, on the fuel rods and inside of the central (unheated) rod. Two pyrometers and a video camera are located at three windows in the upper part of the bundle. After the experiments the post-test examination of the bundle and aerosol samples is carried out with several techniques, including metallography, SEM, microprobe analysis, X-ray radiography and mass spectrometry.
Facility is not 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 Institute for Nuclear Research Pitesti (ICN) has as main activity objective the scientific research, the fundamental and applied technological development, the exploitation of its own research through technology transfer, design, investments, consultancy, expertise and technical specialized assistance, subordinated to ensuring the scientific and technical support for Romania's nuclear energy sector.

HEVA programme started a series of fission-product release experiments, which have been continued with the VERCORS programme. HEVA and the three VERCORS (RT and HT) series represent a total of 28 tests. The results of all these experiments (carried out by CEA with the support of IRSN and EDF) and their interpretation (by IRSN) have made a considerable contribution to the overall knowledge of fission-product release in severe-accident conditions.
Under eight HEVA tests, conducted between 1983 and 1989 where, maximum temperatures were restricted to 2100K. The main objective was to generate information on release of volatile FPs from UO2 fuel. Between 1989 and 1996, the tests VERCORS 1 to 6 were conducted with an improved apparatus and fuel temperatures up to 2600K, so as to expand assessment of release to lower volatility FPs and begin to address issues of UO2 fuel degradation.
Facility is dismanteled.

Facility is not operating.

SOURCE TERM is an international research programme carried out by IRSN and CEA (Commissariat à l’Energie Atomique) with the support of Electricité de France, the European Commission, the US Nuclear Regulatory Commission (US), GDF/ SUEZ/ Tractebel (Belgium), Atomic Energy Canada Limited (Canada), the Paul Scherrer Institute (Switzerland) and the Korea Institute of Nuclear Safety (representing a South Korean consortium).

This programme has a budget of about €30 million over 5 years to investigate four different experimental topics:

1. Studying iodine chemistry.
2. Degradation of boron carbide (B4C) control rods.
3. Consequences of fuel rod heating in air.
4. Fission product releases from irradiated fuel at high temperature.

Facility is closed.

Characterisation methods:

Gaseous releases are continuously measured by a mass spectrometer connected to the furnace outlet. After each test, the different samples are subject to metallographic examination to determine the degradation mechanism : interactions between materials, formation and liquefaction of mixtures, and formation of oxide layers. These examinations are performed with an electron microprobe. As part of tests conducted in the Intermezzo furnace, non-destructive examinations (radiography and tomography) are also performed on control rod sections to characterise their state of degradation.

Facility is not operating.

Mozart program to determine the air-oxidation kinetics of zirconium alloys at high temperature.



Because the current state of knowledge concerning the oxidation in air of zirconium alloy revealed unacceptably large gaps and large uncertainties, IRSN decided to launch the Mozart experimental program to supplement the databases and help to better understand and better model the mechanisms involved.



The temperature range studied was restricted to 600-1200°C, because beyond this temperature range, oxidation becomes catastrophic and precise knowledge of the phenomenon is not required. Different types of alloy (Zircaloy-4, M5TM and Zirlo) in different initial states (virgin, pre-oxidized, pre-hydrided to simulate the initial state of corroded cladding in the reactor) were studied. In these tests, synthetic air was used as the oxidizing medium, and the increased weight gain of the cladding due to oxidation was continuously recorded using a thermobalance.



For initially virgin cladding in the temperature range 800-1000°C, the experimental data revealed two kinetic regimes during oxidation at a given temperature. During the first phase, corresponding to the formation of a dense protective oxide on the initially bare metal, the oxidation rate falls, approximately following a parabolic function. After the cracking of this dense layer ('breakaway'), the second phase is characterized by a faster oxidation rate, which either remains constant or increases over time. The presence of nitrogen plays an important role in the degradation of the cladding in this accelerated regime, because a self-sustaining nitriding/oxidation mechanism generates the formation of a porous, non-protective oxide and causes creep in the cladding. Above 1000°C, the kinetic regime remains parabolic, and therefore rather slow, provided that there is sufficient oxygen in the oxidizing medium. Otherwise, nitrogen becomes preponderant in the oxidizing medium, and can then diffuse in the cladding and cause nitriding. The pre-oxidation layer can either have a protective effect or quite the opposite: an accelerating effect, according to the pre-oxide thickness and the temperature domain and the alloy concerned.



Based on these results, a new model for the oxidation of Zircaloy by air was incorporated in the Astec computational software package developed by IRSN to evaluate the consequences of a core meltdown accident.

Facility is not operating.

In severe accidents with primary-to-secondary leakages, the retention of fission products in horizontal steam generators is poorly understood. The understanding of fission product deposition in realistic steam generator conditions is needed in realistic release estimates in PSA studies, and to design efficient accident management procedures. This is considered very important because steam generator tube rupture sequences are included in the risk dominant sequences.



Tube dimensions of the HORIZON model steam generator and Loviisa VVER-440 steam generators are approximately same. Thus it can be assumed that experiments give realistic results.



In addition to the steam generator section itself, the HORIZON facility includes a lot of equipment needed for steam and aerosol generation, and for measuring the thermal-hydraulic parameters as well as the aerosols concentrations.



The inlet and outlet chamber aerosol mass concentration is monitored with Tapered Element Oscillating Microbalance (TEOM) on-line mass monitor and the particle size distribution is measured on-line with the Electrical Low Pressure Impactor (ELPI). Aerosol sampling system includes heated sampling lines, two diluters (first diluter in system pressure is computer controlled and uses heated dilution air), pressure reducer and sampling valves. It is possible to change sampling point between inlet and outlet chambers.

Facility is dismanteled.

In case of a hypothetical severe accident in a nuclear LWR (light water reactor), the high radiation fields reached in the reactor containment building due to the release of fission products from the reactor core could induce air radiolysis. The air radiolysis products could, in turn, oxidise gaseous molecular iodine into aerosol–borne iodine–oxygen–nitrogencompounds. Thereby, this reaction involves a change of iodine speciation and a decrease of iodine volatility in the reactor containment atmosphere. Kinetic data were produced within the PARIS project on the air radiolysis products formation and destruction, and on their reaction with molecular iodine, with the objective of developing and validating existing kinetic models.
The Program on Air Radiolysis and Iodine adsorption on Surfaces (PARIS) was therefore initiated in 2002 by IRSN in collaboration with AREVA NP, as part of the research programs performed to improve severe accident modelling and evaluation of subsequent fission product release into the environment with specific objective of measuring:
• the rate and amount of ARP production and destruction,
• rate and extent of radiolytic oxidation of molecular iodine into iodine oxides,
• the effect of the containment structural surfaces, namely decontamination coating (“paint”) and stainless steel, on radiolytic oxidation of I2,
• the effect of silver, representing silver-containing aerosol particles, on radiolytic oxidation of I2.
Important new features of the PARIS project were: (1) more realistic low iodine concentrations, (2) surface to volume ratios of paint, steel and silver surface area to containment volume ratio representative of LWR or PHEBUS containments, (3)higher steam fractions and (4) representative dose rates. The PARIS database, containing about 400 tests, was intended to provide data to develop and validate empirical models, and finally to derive a simplified model for ASTEC Code and other severe accident iodine codes.
Facility is not operating.

The purpose of the SISYPHE (Simulation du Système Phébus Enceinte) facility at Cadarache was to build a 1:1 replica of the Phébus FP experimental containment vessel, assisting the Phébus test interpretation, for all phenomena concerning thermal-hydraulics and fission product behaviour.



All systems of the Phébus FP containment vessel are reproduced, except radiation. Instrumentation improved as compared to Phébus, by special optical instruments.



The objective of the testing foreseen in this vessel was manifold:

• A: Thermal Hydraulics Studies: these tests are simulating phenomena like: condensation on condensers and walls, convection, humidity up to 100%, presence of hot/cold sump, steam condensation on aerosols, diffusiophoresis, and iodine affinity with water. B: Aerosol Behaviour Programme: multicomponent aerosols from POLYR generator, soluble or non-soluble. Study on wall deposits or electrophoresis.
• B: Iodine Programme: studying the presence of molecular gaseous iodine, transfer to surfaces, interactions with paint, re-emission from sump by radiolysis, iodine aerosol interaction, and interface with hydrogen.
• C: Mass Transfer Programme: effect of evaporating or non-evaporating sump on molecular iodine mass transfer (in preparation to FPT2 & FPT3). Oxygen is the simulant for iodine. Ultimate goal: mass-transfer model to predict MT coefficient for oxygen and hence for iodine.

Duration of the programme: post-FPT1, between 1995 and 2003.

Facility is dismanteled, part of Phébus.

Iodine is a fission product of major importance, because volatile species can be formed under severe nuclear reactor accident conditions, and may potentially be released into the environment, leading to significant radiological consequences. The CAIMAN programme was devoted to studying the radiochemistry of iodine in the reactor containment in case of a severe accident occurring in a Pressurised Water Reactor; this is a data base of prime importance for the validation of codes, namely IODE, which is a module of the integral ASTEC (Accident Source Term Evaluation Code) code, jointly developed by the IRSN and the GRS. These computations are generally used to predict the radiological consequences of such an accident. The experimental programme, which ran from 1996 to 2002, concerned eighteen experiments in a facility of intermediate scale (300 dm3), where labelled iodine, 131I, was used to perform -counting. The CAIMAN tests are here analysed, and the main experimental observations and trends are described. For each experiment, IODE computations were performed and compared with experimental results in order to assess the possible weak points of the present modelling and to identify key parameters. Broadly speaking, the gaseous concentrations predicted are quite consistent with the experimental ones; the remaining gaps have been identified.
Facility is not operating.

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.

This emission has an impact on the iodine chemistry (AgI) and on the behavior of aerosols in the reactor primary circuit and in the containment. The presence of control rod material influences the source term potentially present in a PWR containment and likely to in the atmosphere.


In a French 900 PWR the AIC material can make up as much as 2 tons.


The results of these experiments should help in establishing computer models on the AIC source term, part of the ESCADRE (later ASTEC) code system. They should also be used as an experimental input for the AIC injection into VERCORS fuel release experiments.

Facility is not operating.

This programme is dedicated to studying iodine chemistry under thermal non equilibrium (impact of chemical kinetics) in the primary cooling system in the event of a core meltdown accident in a water reactor.

The CHIP programme follows two axes which respectivly aims to:

• Identify physico-chemical elements which may have a reaction with iodine during his transfert from core to containment structure (short transfert duration, fast cooling...) and also identify chemical species which influence presence of volatile iodine;
• Get kinetics datas of the main reactions.

So as to fulfill these objectives, the experimental programme uses more or less complex "phenomenolgy" lines in the form of experimental facilities and a "analytical" line.



The data collected will be used to validate the transport models for iodine in the primary cooling system, which are integrated in the ASTEC software. This software is developed by the DPAM to predict the different types of possible accidents and the related radioactive product releases.



The CHIP programme is run by IRSN/DPAM and is part of the International Source Term Programme co-funded by the CEA, EDF, IRSN, the European Commission, the US Nuclear Regulatory Commission (NRC), the Atomic Energy of Canada Limited (AECL), the Korea Institute of Nuclear Safety the Paul Scherrer Institute and SUEZ-Tractebel over the 2005-2012 period.

Facility is in operation.

Since the core meltdown accident in the Three Mile Island reactor in 1979, a series of experimental safety research programmes has been conducted by a number of international research organisations, including the IRSN, which manages the European SARNET network. Simulation models have been developed to calculate the sequence of events in an accident of this type, evaluate its consequences and assess the efficacy of the various measures that can be implemented to limit its effects.



The PHEBUS FP research programme was launched in 1988 by the IRSN (under the former name IPSN – Institut de Protection et de Sûreté Nucléaire, Nuclear Protection and Safety Institute) in partnership with the European Commission and EDF. France collaborated in this programme with the United States, Canada, Japan, South Korea and Switzerland, and five experiments were conducted between 1993 and 2004, involving approximately 80 persons per year.



The main objective was to reduce the uncertainty in evaluating the release of radioactive products in the event of a core meltdown accident in a pressurised water reactor (PWR). To do this, “global” experiments were conducted, that is, experiments in which all the phenomena were represented, from melting of a fuel assembly to release of fission products and structural materials inside a simulated containment vessel, duplicating as closely as possible the conditions that would apply in an accident of this type. The programme had the simultaneous aim of developing and validating the simulation software used to calculate the progression of the accident. This research was to contribute to improving IRSN crisis management by optimising the activities and procedures that would be implemented in the event of a nuclear accident to protect the population and the environment.

Facility is dismanteled.

The first tests were the HEVA series conducted between 1983 and 1989 where, with maximum temperatures restricted to 2100K, the main objective was to generate information on release of volatile FPs from UO2 fuel. Between 1989 and 1996, the tests VERCORS 1 to 6 were conducted with an improved apparatus and fuel temperatures up to 2600K, i.e., higher than those in HEVA so as to expand assessment of release to lower volatility FPs and begin to address issues of UO2 fuel degradation. The apparatus also allowed, to a limited extent, investigation of FP transport.



Between 1996 and 2002, the test series VERCORS HT and RT were performed where the main intention was to improve the database with respect to transport aspects and FP release during later accident phases, i.e., encompassing highly-degraded fuel states (debris, liquefaction). The effect of other parameters on FP release was also studied, notably the influence of high fuel burn-up (up to 70GWd/tU) and mixed-oxide fuel (MOX). The particularity of the RT series was the simplification of the apparatus downstream of the crucible allowing better quantification of low-level releases. Eight RT tests were performed in a variety of atmospheres, namely pure hydrogen, steam-hydrogen mixtures and one test with a helium-air mixture. The particularity of the HT tests was that they included a well-characterized transport system downstream of the crucible allowing deposition and sampling measurements to be made. This series comprised three complementary tests covering the full range of H2-H2O atmospheres with and without the presence of control-rod elements (Ag-In-Cd-B) in order to assess their effect on FP retention.



All six VERCORS tests, most of the VERCORS RT tests and all three HT tests included low- power re-irradiation of the fuel sample in an experimental reactor for a week to re-create the short half-life FP inventory. This allowed the measurement of lower volatility FPs to be considerably enriched. Furthermore, all except three of these tests were conducted with an intermediate temperature plateau (1500K or 1800K) lasting one hour such that full oxidation of the cladding occurred and early sample liquefaction was precluded.



IRSN has for some time been using its computer codes to analyse the results of the VERCORS tests. These codes are the mechanistic release code MFPR [2,3], the simplified release code ELSA [4] (a module of the integral code ASTEC [5]) and the transport code SOPHAEROS [6] (also a module of ASTEC). Regarding the study of FP release, the focus of this work has been on VERCORS 4 and 5 as well as some of the RT tests. In particular, the detailed interpretation made possible with the MFPR code is shedding light on how the oxidation state of the fuel affects the FP chemistry modifying the different fuel phases. Most notably, it is currently thought that the FP Mo, often considered a main component of metallic inclusions during reactor operations, acts to a certain extent as an oxygen buffer (by forming MoO2) with respect to other FPs, i.e., playing a key regulatory role in the chemical forms of the FPs, especially Cs, and hence their volatility and release [7]. With respect to FP transport, the HT tests represent the major interest. Currently, full results of the HT1 test have been available for some time; those of HT3 have just recently been finalized while HT2 was performed in 2002 and results are still awaited. Hence, interpretation of the transport results is, as yet, preliminary [8]. In general terms, it is seen that the usual release categories of volatile, semi-volatile and low volatility FPs can be quite misleading during transport since behaviour can change radically as a function of the in-fuel and ex-fuel differences. One particular point of interest concerns the distinctive difference in behaviour between caesium and iodine where iodine seems to exhibit volatile behaviour whatever the conditions while Cs tends towards different levels of intermediate volatility depending on the reactive species present (molybdenum, boron, etc.). Another aspect being investigated is the conditions leading to significant deposition of so-called semi-volatile FPs (Ba, Mo, Ru, etc.) close to their point of release.

Facility is dismanteled, replaced by VERDON.

No description available.

This programme will help to better estimate the quantity of radioactive iodine released during a core meltdown accident taken into account when elaborating specific emergency plans. The programme results will also be used to better define the means and measures required to limit releases into the environment.



A study on ruthenium chemistry – another radiotoxic product – in the reactor containment under accidental conditions was conducted as part of the ISTP. About twenty tests were performed to assess the effect of irradiation on the volatilisation of ruthenium from the sump or deposits on painted containment surfaces. This study provided experimental data used to determine ruthenium quantities released into the environment in the event of an accident.



Various materials can be irradiated so as to determine the impact of the received dose on the variation in certain properties. This application could be used to study the ageing of polymers, greases and other compounds, which would help improve existing computer models and make it possible to make more informed decisions on the reactor life extension for example.

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.

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.

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.

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.

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.

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.