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).
The ELSMOR facility was built in 2022 at SIET (Piacenza, Italy) in the frame of the ELSMOR European Project (Grant Agreement number: 847553 — ELSMOR NFRP-2018) to validate a decay heat removal system for the E-SMR. Based on the passive mechanisms of natural circulation, the system aims to dissipate the reactor decay heat to a water pool, using two heat exchangers: a plate-type one coupling the primary side to the secondary side, and a vertical tube one coupling the secondary side to the water pool.
ELSMOR HORIZON 2020 PROJECT: https://cordis.europa.eu/project/id/847553/results
EU FASTNET PROJECT CALCULATIONS
Test facility was designed in such a way that it enables to conduct experiments at any inclination of the flow channel. Test facility will be also used for demonstrational and educational purposes. Transparency of the channel walls and the use of WMSs offer students great possibilities to get better understanding on the physics behind the flow behavior. Test section is equipped with in-house manufactured WMSs (32 × 32 wires). The sensors are recording the flow at 5000 frames/s. In addition, special channel section was designed and constructed that enables High-Speed Camera and PIV measurements by minimizing the optical distortions.
To study the behavior of the PCCS configuration planned to be used in the ABWR II concept and to gain experimental data for the validation work of MELCOR severe accident code, a scaled down PCCS model was designed and constructed at Lappeenranta University of Technology in Finland in 2012–2013. The PCCS model is connected to the drywell and wetwell compartments of PPOOLEX, which is acting as a host facility. Steam needed in the experiments is produced with the nearby PACTEL facility. The PCCS model consists of five horizontal U-tube shaped heat exchange tubes installed inside a secondary side liquid pool. The pool is in atmospheric pressure and covered by a lid with an exit pipe out of the laboratory and with viewing windows for video cameras.
The PPOOLEX facility consists of a pressure vessel containing a wet well compartment (condensation pool), dry well compartment, inlet plenum and air/steam line piping. An intermediate floor separates the compartments from each other but a route for gas/steam flow from the dry well to the wet well is created by a vertical blowdown pipe attached underneath the floor.
PACTEL is a volumetrically scaled (1: 305) facility including a pressurizer, high and low pressure emergency core cooling systems, and accumulators. The reactor vessel is simulated with a U-tube construction including separate downcomer and core sections. The core itself is consists 144 full-length, electrically heated fuel rod simulators. Component heights and relative elevations correspond to those of the full scale reactor to match the natural circulation gravitational heads in the reference system. Three coolant loops with double capacity steam generators are used to model six loops of the reference power plant. The facility is still in operation for example as an auxiliary system for the separate effect test facilities. Until now, 239 experiments have been carried out with the facility.
The PWR PACTEL facility consists of a reactor pressure vessel model, two loops with vertical steam generators, a pressurizer, and emergency core cooling systems. The new loops and steam generators of EPR style construction enable the PWR and EPR related experimental research. The pressure vessel model in PWR PACTEL comprises a U-tube construction modeling the downcomer, lower plenum, core and upper plenum. The core rod bundle consists of 144 electrically heated fuel rod simulators arranged in three parallel channels. The core can be powered by a maximum of 1 MW electric power supply. The maximum core power corresponds roughly to the scaled residual heating power of the EPR reactor. The total height of the PWR PACTEL pressure vessel model corresponds to the pressure vessel height of EPR. The volume ratio between the pressure vessels in PWR PACTEL and EPR is about 1/405.
The LOBI Project originated from a reactor safety research and development contract between the European Commission and the former Bundesminister für Forschung und Technologie of the Federal Republic of Germany. On the basis of contingent and perceived safety requirements, BMFT LOBI Project decided in 1972 on the need of an experimental programme to be conducted in a integral system test facility to investigate thermal-hydraulic phenomenologies relevant to accident conditions in pressurized water reactors (PWRs) of German design. As result of a tender, the execution of this study was awarded to the Joint Research Centre of the European Commission.
Objectives:
- Investigation of Basic Phenomenologies governing the thermal-hydraulic response of an Integral System Test Facility for a range of PWR Operational and Accident Conditions.
- Provision of an Experimental Data Base for the Development and Improvement of Analytical Models and the Independent Assessment of Large System Codes used in LWR Safety Analysis.
Facility Configurations
The LOBI-MOD1 test facility configuration was designed to meet the relevant requirements originating from the objectives of the experimental investigations of Large and Intermediate Break Loss of Coolant Accidents (LOCA). A total of 28 tests were performed with this configuration during the period December 1979 to June 1982.
The LOBI-MOD2 test facility configuration, operating since April 1984, represents an upgraded version designed to meet also all relevant requierements related to the investigation of Small Break LOCAs and Special Transients. A total of 42 tests were performed in the period April 84 to June 1991.
Facility is dismanteled.
Test Matrix
In the PDF files attached below you can find more information on the Facility and a very useful Test Matrix with information of each experiment.
CERTA-TN
Also attached you may find information of CERTA - TN: European Thematic Network for the Consolidation of the Integral System Effect Experimental Databases for Reactor Thermal-Hydraulic Safety Analysis in which JRC participated with the LOBI Facility data and that was the origin to the development of the first version of the STRESA Database.
The activities covered the following three areas:
- Thermal hydraulic calculations described the cooling conditions possibly established during the incident.
- Simulation of fuel behaviour described the oxidation and degradation mechanisms of fuel assemblies.
- The release of fission products from the failed fuel rods was estimated and compared to available measured data.
The produced numerical results improved the understanding of the causes and mechanisms of fuel failures during the Paks-2 incident and provided new information on the behaviour of nuclear fuel under accident conditions.
No description available.
ARIGS is one of the programs on the aerosol retention on the tubes surrounding the breach within the secondary side of the steam generator in the absence of water. Its development has been internationally framed within the EU-SGTR and the ARTIST programs. Experimental activities were focused on setting up a reliable database in which the influence of gas mass flow rate, breach configuration and particle nature in the aerosol retention are properly considered. Theoretical activities are aimed at developing a predictive tool (ARISG) capable of assessing source term attenuation in the scenario with reasonable accuracy. Given the major importance of jet aerodynamics, 3D CFD analyses are being conducted to assist both test interpretation and model development.
ARISG-I was a step forward in the modeling of the aerosol retention of the steam generator. According to this analysis the main areas where research is needed are: gas jet behavior across the tube bank; particle resuspension, erosion, and/or bouncing; and particle inertial impaction and turbulent deposition under foreseen conditions.
The RUSET experimental programme was launched in 2002 at the Hungarian Academy of Sciences KFKI Atomic Energy Research Institute (AEKI). The aim of the program was to get data for assessment of ruthenium release at severe accident with air ingress. More than forty small scale tests have been performed with mixed powder components of inactive materials and with short fuel rods. The influence of temperature, air flow rate and the presence of other fission products on the gaseous Ru release and the retention role of fuel pellets and cladding have been investigated. The test series indicated that if an air ingress type severe accident occurs most of the initial Ru mass can be released from the reactor core to the containment or environment. Some part of the released gaseous Ru undergoes precipitation and deposits on the cold surfaces, another part is released in gaseous form. The deposited Ru oxides can serve as a secondary source for further gaseous Ru release.
Since the beginning of the 1990s’, several experimental series have been performed at the Hungarian Academy of Sciences KFKI Atomic Energy Research Institute (AEKI) with E110 and Zircaloy claddings. The aims of these experiments were to study and to compare the mechanical properties of the cladding materials and to investigate the effect of oxidation and hydrogen uptake on the mechanical performance of the claddings under accident conditions. The objectives have been achieved through separate effect tests with well defined conditions.
The main results of the tests have been collected into the Experimental Database of E110 Claddings under Accident Conditions. The database involves the data of several experimental series in the following main directories:
- Cladding ballooning tests (more than 170 tube tests).
- Tensile tests with tube and sheet specimens (more than 100 samples).
- Oxidation tests between 500-1200 °C.
- Ring compression tests (more than 100 samples).
Most of the tests were carried out with E110 cladding, but several experimental points were produced with Zircaloy-4 cladding as well for comparison purposes. The database includes not only the on-line measured data, but the results of post-test examinations (visual observations, metallographic analysis, SEM analysis). Experimental technical reports and some selected papers describing the tests are also available in the database.
The high temperature interaction of reactor core materials during a severe accident leads to the oxidation and melting of metal components. The interactions of molten Zr cladding with uranium-dioxide and zirconium-dioxide are important factors in the determination of fuel failure conditions and play a role in the loss of fuel rod like geometry and in the formation of debris bed and molten pool in the core.
The dissolution of uranium-dioxide and zirconium-dioxide by molten Zircaloy were investigated in earlier separate effect tests. UO2 dissolution experiments were carried out with UO2 crucible and Zr charge. The simultaneous UO2-ZrO2 molten Zircaloy dissolution was investigated in UO2 crucibles with Zircaloy charge and ZrO2 central rod [1]. The analysis of different experiments showed some discrepancy between the results, which was connected with different crucible sizes, UO2/Zr mass ratios and the melt surface to volume ratios [2]. The comparison of simultaneous dissolution tests of UO2/ZrO2 and separate UO2 and ZrO2 showed faster dissolution and larger extent of dissolution in the case of simultaneous tests. These observations emphasized the importance of prototypic conditions on the dissolution process. For these reasons experiments with short fuel rod segments were carried out in the KFKI Atomic Energy Research Institute, Budapest. The development of dissolution models needs more prototypical experiments, this task was addressed in the COLOSS project of the 5th FWP. The analytical support of the AEKI short fuel rod dissolution tests was provided by COLOSS partners [3],[4]. The results of the experimental series are expected to make possible the further model development and code validation. 2. OBJECTIVES
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.
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:
- Studying iodine chemistry.
- Degradation of boron carbide (B4C) control rods.
- Consequences of fuel rod heating in air.
- Fission product releases from irradiated fuel at high temperature.
Facility is closed.
Facility is not operating.
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.
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.
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.
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.
No description available.
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.
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.
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.
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.
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.
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.
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.
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.
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.
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.
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 catalyst sheets (stainless steel coated with washcoat/platinum catalyst material) are arranged in parallel forming vertical rectangular flow channels. Such a set-up represents a box-type recombiner section of AREVA design. Inside of the configuration the distribution of the catalyst temperatures and the gas com- position in vertical flow direction are measured. The correlation of the hydrogen conversion and catalyst temperatures with the experimental parameters serve basically to clarify the interactions of reaction kinetics, heat and mass transfer, and the flow conditions inside the recombiner.
Facility is in operation.
The 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.
TUBA (thermo-phoresis and diffusiophoresis) programme included experiments represented conditions expected in the SG tubes. Laminar flow was used in TUBA tests (TUBA-T: thermophoresis and TUBA-D: diffusiophoresis & thermo-diffusiophoresis).
From the comparison of experimental results and SOPHAEROS calculations it was concluded that there was sufficient agreement for most of the cases studied.
Facility is dismanteled.
RECI is a 2½ year experimental programme that was brought to completion as of October 2004. The aim of the RECI (RECombiner & Iodine) program was to quantify the iodide → iodine conversion in realistic conditions of recombiner operation, albeit under the following constraints: the experiments were to be performed with non-radioactive substances, and without hydrogen. The comprehensive tests grid allowed to investigate into the decomposition of cesium and cadmium iodides under thermal-hydraulics conditions that mimics the recombiner operation, despite the technical limitations of the RECI test bench.
The aerosol generator selected limits the RECI programme to the study of water soluble substances, namely cesium and cadmium iodides: silver iodide is insoluble in water, and indium monoiodide is hydrolysed. However, the experimental results can be interpolated with reasonable confidence, since CsI and CdI2 are the two end-terms in the stability range of the relevant iodides. The instability of metal iodides, in a wet and oxidizing atmosphere, already demonstrated in chemistry laboratories, has been confirmed in more relevant physico-chemical conditions. The high conversion yields obtained do not come as a surprise since the RECI experiments provide a close analogy to the processes known as spray drying and spray (reactive-, or oxidizing-) pyrolysis, widely used in the laboratory and in the manufacturing industry. Both processes capitalize upon the high surface/volume ratio of aerosol particles to master comparatively slow chemical reactions and to produce nano-particles, the precursor material being often a finely powdered metal halide.
The experimental test bench consists of 4 units.
- Aerosol generation: an ultrasonic aerosol generator atomises the aqueous solution of a water soluble iodide. Monodispersed droplets are then dried, yielding iodide particles, the size of which is determined by the concentration of the solution. The input power of the ultrasonic acoustic transducer sets the aerosol concentration.
- Recombiner surrogate: a clear fused quartz or alumina tubing, which can accommodate a catalyst foil, is heated in a vertical tube furnace.
- Aerosol characterization: particles concentration & size distribution measurements. • Gaseous iodine analysis: 3 independently calibrated methods are implemented in the flue gas.
Facility is dismanteled.
In order to comply with experimental device design requirements, different devices were developped, tested and set up on CARAIDAS:
- experimental enclosure in which representative thermodynamic conditions could be achieved,
- the monosized drops generator, the drops diameter measurements and the drops collector,
- the cesium iodide aerosols generator, concentration and size distribution measurements.
Facility is not operating.
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.
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 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.
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.