The atmospheric chemistry of iodine is important in different environments, and particularly in case of a nuclear reactor severe accident. This paper describes the performance of an Incoherent Broad-Band Cavity Enhanced Absorption Spectroscopy (IBB-CEAS) device, based on a green LED source allowing the online detection of gaseous molecular iodine. High sensitivities and a wide dynamic range are obtained. The performances and limitations of the IBB-CEAS technique for measurements of iodine in the presence of aerosols and water vapour are discussed. The capabilities of this device to perform kinetics studies on iodine have been investigated in the framework of nuclear reactor accident research. The re-vaporization process of deposited CsI by optical monitoring of I-2 concentration-time profiles is presented. The aerosols used in the experiments have been well characterized (size, distribution, concentration and mass) before interaction with iodine by a NanoScan scanning mobility particle sizer (SMPS), a PEGASOR particle sensor PPS-M and a tapered element oscillating microbalance (TEOM). (C) 2018 Elsevier Ltd. All rights reserved.

The PASSAM (Passive and Active Systems on Severe Accident source term Mitigation) project was launched in the frame of the 7th framework programme of the European Commission. Coordinated by IRSN, this four year project (2013-2016) involved nine partners from six countries: IRSN, EDF and university of Lorraine (France); CIEMAT and CSIC (Spain); PSI (Switzerland); RSE (Italy); VTT (Finland) and AREVA GmbH (Germany). It was mainly an R&D project of experimental nature aimed at investigating phenomena that might enhance source term mitigation in case of a severe accident in a Nuclear Power Plant (NPP). Both existing systems (i.e., water scrubbing and sand bed filters plus metallic pre-filters) and innovative ones (i.e., high pressure sprays, electrostatic precipitators, acoustic agglomerators and, advanced zeolites and combined wet-dry filtration systems), were experimentally studied in conditions as close as possible from those anticipated for severe accidents. This paper presents the main experimental results of the project which represent a significant extension of the current database on these existing or innovative mitigation systems. Application of some of these data for improving existing models or developing new ones should eventually enhance the capability of modelling Severe Accident Management measures and developing improved guidelines. (C) 2018 Elsevier Ltd. All rights reserved.

The severe accident integral code ASTEC, jointly developed since almost 20 years by IRSN and GRS, simulates the behaviour of a whole nuclear power plant under severe accident conditions, including severe accident management by engineering systems and procedures. Since 2004, the ASTEC code is progressively becoming the reference European severe accident integral code through in particular the intensification of research activities carried out in the frame of the SARNET European network of excellence.The first version of the new series ASTEC V2 was released in 2009 to about 30 organizations worldwide and in particular to SARNET partners. With respect to the previous V1 series, this new V2 series includes advanced core degradation models (issued from the ICARE2 IRSN mechanistic code) and necessary extensions to be applicable to Gen. III reactor designs, notably a description of the core catcher component to simulate severe accidents transients applied to the EPR reactor. Besides these two key-evolutions, most of the other physical modules have also been improved and ASTEC V2 is now coupled to the SUNSET statistical tool to make easier the uncertainty and sensitivity analyses. The ASTEC models are today at the state of the art (in particular fission product models with respect to source term evaluation), except for quenching of a severely damage core.Beyond the need to develop an adequate model for the reflooding of a degraded core, the main other mean-term objectives are to further progress on the on-going extension of the scope of application to BWR and CANDU reactors, to spent fuel pool accidents as well as to accidents in both the ITER Fusion facility and Gen. IV reactors (in priority on sodium-cooled fast reactors) while making ASTEC evolving towards a severe accident simulator constitutes the main long-term objective.This paper presents the status of the ASTEC V2 versions, focussing on the description of V2.0 models for water-cooled nuclear plants. (C) 2013 Elsevier B.V. All rights reserved.

One of the most disputed issues raised by molten corium concrete interaction (MCCI) is how the 2D cavity ablation in an oxidic pool evolves: why is the ablation anisotropic with siliceous concretes and isotropic with carbonaceous concretes. The work performed in the frame of the SARNET2 WP6 group during the last 4 years has enabled significant progress on this topic. This paper summarizes this progress using the analysis of recent 2D real material experiments in an oxidic pool and from analytical simulant experiments on 2D heat convection in a bubbling pool, including calculations and recalculations with MCCI codes available in Europe.Firstly, the effective heat transfer coefficients from the bulk pool to the bottom and lateral pool interfaces deduced from MCCI experiments lead to a range of a few 100 W/m(2)/K. By contrast, a detailed review of possible 2D convection mechanisms shows that the individual heat convection mechanisms (without taking a crust into account) such as gas bubbling convection and solutal convection overestimates the overall heat transfer coefficient, and does not account for the main trends of 2D ablation deduced from MCCI tests, which are very dependent on the composition of concrete components and aggregates. This fact, in turn, points to the effect of more complex pool/concrete interface structures. On the basis of a thorough interpretation of the experimental database and of a detailed comparison of MCCI code predictions, a set of the most realistic and consistent assumptions are identified and major remaining uncertainties are listed. (C) 2014 Elsevier Ltd. All rights reserved.

The 5 FWP EURSAFE project highlighted iodine chemistry in the containment as one of the issues requiring further research in order to reduce source term uncertainties. Consequently, a series of studies was launched in the 6 FWP SARNET project aimed at improving the predictability of iodine behaviour during severe accidents via a better understanding of the complex chemical phenomena in the containment. In particular, SARNET has striven to foster common interpretation of integral and separate effect test data; production of new or improved models where necessary needed, and compilation of the existing knowledge of the subject. The work has been based on a substantial amount of experimental information made available from bench-scale projects (PARIS and EPICUR), via intermediate-scale tests (CAIMAIN) to large scale facilities (SISYPHE, THAI and PHEBUS-FP).

In a hypothetical severe accident in a Pressurised Water Reactor (PWR), Fission Products (FPs) can be released from the overheated nuclear fuel and partially transported by gases, composed of a mixture of superheated steam and hydrogen, to the reactor containment. Subsequent air ingress into a damaged reactor core may lead to enhanced fuel oxidation, affecting some FP release, especially that of ruthenium. Ruthenium is of particular interest because of its high radiotoxicity and its ability to form very volatile oxides. In the reactor containment, such volatile forms are very hazardous as they are much less efficiently trapped than particulate forms by emergency filtered venting.In the four and a half years of SARNET, collaborative research dedicated to the "ruthenium story" has been performed by several partners. This paper presents the main achievements over the whole project period.Starting from experimental observations showing that fuel could be extensively oxidised by air to, and that a significant fraction of ruthenium inventory can be released, rather satisfactory models have been developed. In addition, the effect of the air interaction with Zircaloy cladding, as well as with UO2 itself, has been studied.Experiments on the complex transformations of ruthenium oxides upon cooling through the reactor circuit have been performed. An unexpectedly large effect of temperature on the decomposition rate of gaseous ruthenium compounds has been found. as well as effects of the nature of circuit internal surfaces and other FP deposits. So it has been highlighted that various forms of ruthenium can reach the containment, but the most probable gaseous species under these conditions is ruthenium tetroxide. Preliminary analysis of ruthenium transport supports these conclusions.Experiments and analysis have also been launched on the radio-chemical reactions undergone by these ruthenium oxides in the reactor containment. Competing effects of gaseous decomposition to solid particles and re-volatilization from these ruthenium deposits have been demonstrated and modelled.The paper concludes by identifying the remaining work needed to achieve full resolution of the ruthenium Source term issue. Recommendations are made for future research activities in the follow-up programme SARNET2. (C) 2009 Elsevier Ltd. All rights reserved.

Summary During a hypothetical severe accident on a nuclear Pressurized Water Reactor (PWR), the formation of highly radiotoxic RuO4(g) may occur in the reactor containment building, resulting from the interactions of ruthenium oxide deposits with the oxidising medium induced by air radiolysis. Consequently, there is a risk that the gaseous ruthenium tetroxide could be dispersed into the environment through containment leakages; therefore data concerning the behaviour of ruthenium oxide deposits is of primary importance for safety studies. An experiment has been designed to study the interactions of ruthenium oxide deposits with ozone, formed by air radiolysis. Experimental results have shown that the oxidation reaction leading to the formation of RuO4(g) occurred to a large extent. An oxidation kinetic rate law has been determined, in dry and moist air, for ruthenium deposited onto painted substrates, representative of the inner surfaces of PWR's. This law, combined with tetroxide stability data [Mun, C., Cantrel, L., Madic, C.: Study of RuO4 decomposition in dry and moist air. Radiochim. Acta 95(11), 1 (2007)], will allow a first evaluation of ruthenium revolatilisation phenomenon, possibly occurring in the containment during a severe accident.

In the case of a hypothetical severe accident in a nuclear pressurized water reactor, the formation of radiotoxic RuO4(g) may occur in the reactor containment building, resulting from the interactions of ruthenium oxide deposits with the oxidizing medium induced by air radiolysis. Consequently, this gaseous ruthenium tetroxide may be dispersed into the environment; therefore, the determination of the ruthenium deposits behavior is of primary importance for nuclear safety studies. An experimental study, performed by the French Institut de Radioprotection et de Surete Nucleaire (IRSN), using a gamma irradiator cell (EPICUR facility at IRSN/Cadarache) has been carried out in order to obtain experimental data on these interactions. The results showed that radiolytic oxidation of ruthenium oxide deposits leads to the formation of gaseous ruthenium tetroxide to a significant extent. A comparison between the revolatilized Ru fractions obtained experimentally and those obtained by calculations based on the rate laws modeling ozone irradiation effect, established in previous studies, is presented. The disagreement observed is discussed. It appears that the oxidation resulting from air/steam radiolysis products is enhanced in comparison with pure ozone effect.

In the case of a hypothetical severe accident in a nuclear power plant, interactions of gaseous RuO4 with reactor containment building surfaces (stainless steel and epoxy paint) could possibly lead to a black Ru-containing deposit on these surfaces. Some scenarios include the possibility of formation of highly radiotoxic RuO4(g) by the interactions of these deposits with the oxidizing medium induced by air radiolysis, in the reactor containment building, and consequently dispersion of this species. Therefore, the accurate determination of the chemical nature of ruthenium in the deposits is of the high importance for safety studies. An experiment was designed to model the interactions of RuO4(g) with samples of stainless steel and of steel covered with epoxy paint. Then, these deposits have been carefully characterised by scanning electron microscopy (SEM/EDS), electron probe microanalysis (EPMA) and X-ray photoelectron spectroscopy (XPS). The analysis by XPS of Ru deposits formed by interaction of RuO4(g), revealed that the ruthenium is likely to be in the IV oxidation state, as the shapes of the Ru 3d core levels are very similar with those observed on the RuO2·xH2O reference powder sample. The analysis of O 1s peaks indicates a large component attributed to the hydroxyl functional groups. From these results, it was concluded that Ru was present on the surface of the deposits as an oxyhydroxide of Ru(IV). It has also to be pointed out that the presence of “pure” RuO2, or of a thin layer of RuO3 or Ru2O5, coming from the decomposition of RuO4 on the surface of samples of stainless steel and epoxy paint, could be ruled out. These findings will be used for further investigations of the possible revolatilisation phenomena induced by ozone.

During a hypothetical severe nuclear accident on a pressurized water reactor (PWR), it is of primary importance to assess potential radionuclide release into the environment, and thus to better understand the volatile ruthenium tetroxide stability, in the containment building, due to its high radiotoxicity. The stability of RuO 4 (g) in dry and moist air, under conditions representative of a PWR containment building, is investigated. RuO 4 decomposition occurs in bulk gas phase, without any specific affinity with surfaces. The kinetic rate law of RuO 4 reduction is found to be dependent on the presence of steam. The humidity seems to play a catalytic role, as well as the presence of ruthenium dioxide deposits. The temperature is also a key parameter. In the presence of steam, the half-life times of RuO [4] are found to be respectively of 5 h and 9 h at 90 °C and 40 °C. A chemical reaction scheme consistent with the experimental observations is proposed.

During a hypothetical severe accident in a pressurized water reactor (PWR), fission products (FPs) are released from the nuclear fuel and may reach the reactor containment building. Among the FPs, ruthenium is of particular interest due to its ability to form volatile oxide compounds in highly oxidizing conditions. In addition, ruthenium is a very hazardous compound because it is chemically toxic and also because of its radiotoxicity.The topic of ruthenium is examined in terms of nuclear safety issues. A review of the literature regarding ruthenium oxides properties, gaseous and aqueous chemistry is compiled. The study focuses on ruthenium tetroxide (RuO4), which is highly reactive and volatile and is the most likely gaseous chemical form under the conditions prevailing in the containment. The interactions between ruthenium oxides and containment surfaces, which could be most important in overall ruthenium behavior, are also discussed. Finally, an evaluation of the possible revolatilization phenomena of ruthenium adsorbed on PWR containment surfaces or dissolved in the sump under superoxidizing conditions (radiolysis) is also presented. In this case, ruthenium dioxide (RuO2) must also be considered.Knowledge of all these phenomena is required to accurately predict ruthenium behavior and to make a best-estimate assessment of the potential ruthenium source term.