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International audience; Within the NURESAFE project, a main steam line break benchmark has been defined and solved by codes integrated into the European code platform NURESIM. The paper describes the results of the calculations for this benchmark. Six different solutions using different codes and code systems are provided for the comparison. The quantitative differences in the results are dominated by the differences in the secondary system parameters during the depressurization. The source of these differences comes mainly from the application of different models for the two-phase leak flow available in the system codes. The use of two different thermal hydraulic system codes influences the results more than expected when the benchmark was created. The codes integrated into the NURESIM platform showed their applicability to a challenging transient like a main steam line break. © 2017 Elsevier B.V.

International audience; The international OECD/NEA FIRE (Fire Incidents Records Exchange) Database is one of the nuclear power plant (NPP) operational events Database Projects currently operated under the auspices of the Organisation for Economic Co-operation and Development (OECD) Nuclear Energy Agency (NEA). This Database for the collection of detailed information from fire events at NPPs from fourteen NEA member countries is approaching the end of its fifth phase. The FIRE Database is considered mature enough for applications, with respect to not only deter¬ministic safety assessment but also in fire probabilistic risk assessment.The most recent version of this Database covers more than 500 well documented fire events during all operational phases of the plant life cycle from construction up to the longer duration safe shutdown, as well as a few events from NPP units under decommissioning. The number of recorded events increases continuously within each annual update. The Database structure enables analysts to carry out search queries for different aspects of fire events and allows investigations into even more complex fire scenarios. Various analyses can be systematically performed in an automated manner, from generating different samples up to a more or less complete statistical analysis.This paper presents a brief overview of the application possibilities of the OECD/NEA FIRE Database for supporting NPP operators as well as regulators in assessing fire safety issues.

International audience; In-Vessel Retention (IVR) of corium is one of the possible strategies for Severe Accident (SA) mitigation. Its main advantage lies in the fact that, by maintaining the corium within the vessel, the integrity of the last containment barrier against corium aggression is preserved. One of the issues for the demonstration of the success of this strategy is the evaluation of the behaviour of the corium relocated in the lower head and how it stabilizes and affects the integrity of the vessel wall. The first modelling was developed in the nineties and assessed the heat transfers in a stratified corium pool with a top metal layer made of steel and Zirconium only. About 10 years later, the results of the MASCA program highlighted the possibility of having more complex stratified configurations, including a dense metal layer. Addressing thermochemical effects on the stratification makes the modelling of the corium pool in the lower head more difficult and, in addition, knowledge of the associated kinetics is still limited. As a consequence, available SA codes, either integral or dedicated to the lower head, can differ significantly in their models, which leads to discrepancies in the results when evaluating the IVR strategy.In order to identify the main modelling issues and to assess the capabilities of the codes, a benchmark exercise for code validation was made in the scope of the European H2020 project IVMR (In-Vessel Melt Retention). It is based on the definition of different IVR configurations at reactor scale with an increase in the complexity of the phenomena involved: starting from a steady-state stratified pool with metal on the top, up to consideration of corium phase separation at thermochemical equilibrium and progressive ablation of metallic structures and vessel wall. In this paper, the main results and outcomes obtained are presented and discussed. Six organisations took part in this benchmark (CEA, EDF, GRS, IBRAE, IRSN, NRC-KI) and 6 different codes were used (ASTEC, ATHLET-CD, MAAP_EDF, PROCOR, HEFEST_URAN and HEFEST – stand-alone version of the corresponding module of the SOCRAT code). Finally, sensitivity studies are performed and allow obtaining a more consolidated range of results.Thanks to this benchmark exercise and to the approach followed with a progressive increase of complexity, the capabilities of codes to evaluate the heat flux profile applied to the vessel wall in steady-state are demonstrated.Then, the larger differences between code results obtained in transient situations have been identified and associated modelling assumptions discussed.

This paper presents the experimental characterization of a spray produced by a VVER-440 nuclear reactor type nozzle. Several droplet size and velocity profiles have been obtained at different pressure supplies and different heights below the outlet of the spray nozzle. Repeatability and stability have been checked. A log-normal size distribution can be fitted on the experimental results. Correlations between droplet velocities and sizes at different locations are also given, showing that for small droplet sizes (below 300 μm) no clear size-velocity correlation exists below 0.7 m from nozzle outlet, but for larger droplets, a classical evolution of this correlation is observed. It is concluded that the experimental data obtained at 300 mm from the nozzle outlet can be used as spray boundary conditions for numerical calculations with CFD codes. The other experimental data (at 500 and 700 mm from the nozzle outlet) can serve for detailed code validation, if the correlations between sizes and velocities are considered in the validation procedure: indeed, the averaging of droplet sizes and velocities can mask some typical spray results on droplet sizes and velocities evolutions. If a good code validation of the size-velocity correlations is obtained at 500 and 700 mm from the nozzle outlet, the concerned code may then be used with good confidence to extrapolate the results at other distances (for example, 3 m, 5 m) which cannot be obtained easily experimentally. © 2016 by Begell House, Inc.

International audience; The In-Vessel Retention (IVR) strategy for Light Water Reactors (LWR) intends to stabilize and isolate corium and fission products in the reactor pressure vessel and in the primary circuit. This type of Severe Accident Management (SAM) strategy has already been incorporated in the SAM guidance (SAMG) of several operating small size LWR (reactor below 500 MWe (like VVER440)) and is part of the SAMG strategies for some Gen III + PWRs of higher power like the AP1000 or the APR1400. However, for high power reactors, estimations using current level of conservatism show that RPV failure caused by thermo-mechanical rupture takes place in some cases. A better estimation of the residual risk (probability of cases with vessel rupture) requires the use of models with a lower level of conservatism. In Europe, the IVMR project aims at providing new experimental data and a harmonized methodology for IVR. A synthesis of the methodology applied to demonstrate the efficiency of IVR strategy for VVER-440 in Europe (Finland, Slovakia, Hungary and Czech Republic) was made. It showed very consistent results, following quite comparable methodologies. The main weakness of the demonstration was identified in the evaluation of the heat flux that could be reached in transient situations, e.g. under the “3-layers” configuration, where the “focusing effect” may cause higher heat fluxes than in steady-state (due to transient “thin” metal layer on top). Analyses of various designs of reactors with a power between 900 and 1300 MWe were also made. Different models for the description of the molten pool were used homogeneous, stratified with fixed configuration, stratified with evolving configuration. The last type of model provides the highest heat fluxes (above 3 MW/m2) whereas the first type provides the lowest heat fluxes (around 500 MW/m2) but is not realistic due to the non-miscibility of steel with UO2. Obviously, there is a need to reach a consensus about best estimate practices for IVR assessment to be used in the major codes for safety analysis, such as ASTEC, MELCOR, SOCRAT, MAAP, ATHLET-CD, SCDAP/RELAP, etc. Despite the model discrepancies, and leaving aside the unrealistic case of homogeneous pool, the average calculated heat fluxes in many cases are well above 1 MW/m2 which could reduce the residual thickness of the vessel considerably and threaten its integrity. Therefore, it is clear that the safety demonstration of IVR for high power reactors requires a more careful evaluation of the situations which can lead to formation of either a very thin top metal layer provoking focusing effect or significantly overheated metal, e.g. after oxide and metal layer inversion. It also requires an accurate mechanical analysis of the ablated vessel. The current approach followed by most experts for IVR is a compromise between a deterministic analysis using the significant knowledge gained during the last two decades and a probabilistic analysis to take into account large uncertainties due to the lack of data for some physical phenomena (such as transient effects) and due to excessive simplifications of models. A harmonization of the positions of safety authorities on the IVR strategy is necessary to allow decision making based on shared scientific knowledge. Currently, the acceptance criteria of a safety demonstration for IVR may be differently defined from one country to the other and the differences should be further discussed to reach harmonization on this important topic. This includes the accident scenarios to be considered in the demonstration and the modelling of the phenomena in the vessel. Such harmonization is one of the goals of IVMR project. A revised methodology is proposed, where the safety criterion is based not only on a comparison of the heat flux and the Critical Heat Flux (CHF) profiles as in current approaches but also on the minimum vessel thickness reached after ablation and the maximum integral loads that is applied to the vessel during the transient. The main advantage of this revised criterion is in consideration of both steady-state and transient loads on the RPV. Another advantage is that this criterion may be used in both probabilistic and deterministic approaches, whereas the current approaches are mostly deterministic (with deterministic calculations used only for estimates of uncertainty ranges of input parameters). © 2018 Elsevier Ltd

International audience; In a severe accident transient, corium composition and its properties determine its behaviour and its potential interactions both with the reactor vessel and in the later phases with the concrete basemat. This, in turn, requires a detailed knowledge of the phases present at temperature and how they are formed. Because it implies mainly the investigation of chemical systems at high temperature, these data are often difficult to obtain or are uncertain if it already exists. Therefore more data are required both to complete the thermodynamic databanks (such as NUCLEA) and to construct accurate equilibrium phase diagrams and to finally contribute to the improvement of the codes simulating these severe accident conditions. The MCCI work package (WP6) of the SARNET 2 Network of Excellence has been addressing these problems. In this framework in large facilities such as VULCANO tests have been performed on the interactions and ablation of UO2-containing melts with concrete. They have been completed by large scale MCCI testing such EPICOR on vessel steel corrosion. In parallel in major EU-funded ISTC projects co-ordinated with national institutes, such as the CORPHAD and PRECOS, smaller, single effect tests have been carried out on the more difficult phase diagrams. These have produced data that can be directly used by databanks and for modelling improvement/validation. From these data significant advances in the melt chemistry and pool behaviour have been made. A selection of experiments from participating institutes are presented in this paper and give hindsight into the major processes and so give clear indications for the future work, especially in light of the Fukushima accident. © 2014 The Authors.

International audience; No individual country has sufficient resources to address all severe accident important phenomena within the framework of a national research programme, therefore optimised use of resources and the collaboration at European and international level are very important. Integrating European severe accident research facilities into a pan-European laboratory for severe accident and corium studies and providing resources to other European partners for better understanding of possible accident scenarios and phenomena is necessary in order to improve safety of existing and, in the long-term, of future reactors.SAFEST (Severe Accident Facilities for European Safety Targets) is a European project networking the European corium experimental laboratories with the objective to establish coordination activities, enabling the development of a common vision and of research roadmaps for the next years, and of the management structure to achieve these goals. One of the main objectives is to address and resolve the variety of the remaining severe accident issues related to accident analysis and corium behaviour. The project is a valuable asset for the fulfilment of the severe accident RetD programmes that are being set up after the Fukushima Daiichi accidents and the subsequent European stress tests, addressing both national and European objectives.Roadmaps on European severe accident experimental research for water reactors and for GenIV technologies will be drafted. Joint RetD is conducted to improve the excellence of the SAFEST facilities this includes measurement of corium physical properties, improvement of instrumentation, consensus on scaling law rationales and cross comparison of material analyses.Joint experimental research is a clear objective in the SAFEST project to provide solutions for stabilisation of severe accident and termination of consequences for the current Gen II and III plants. Consequently, the knowledge obtained in SAFEST shall lead to improved severe accident management measures, which are essential for reactor safety. In addition, it will offer competitive advantages for the nuclear industry and contribute to the long-term sustainability of nuclear energy.

International audience; Significant efforts are put into the assessment of the severe accident integral code ASTEC, jointly developed since several years by IRSN and GRS, either through comparison with results of the most important international experiments or through benchmarks with other severe accident simulation codes on plant applications. These efforts are done in first priority by the code developers' organisations, IRSN and GRS, and also by numerous partners, in particular in the frame of the SARNET European network. The first version of the new series ASTEC V2 had been released in July 2009 to SARNET partners. Two subsequent V2.0 code revisions, including several modelling improvements, have been then released to the same partners, respectively in 2010 and 2011. This paper summarises first the approach of ASTEC validation vs. experiments, along with a description of the validation matrix, and presents then a few examples of applications of the ASTEC V2.0-rev1 version carried out in 2011 by the SARNET users. These calculation examples are selected in a way to cover diverse aspects of severe accident phenomenology, i.e. to cover both in-vessel and ex-vessel processes, in order to provide a good picture of the current ASTEC V2 capabilities. Finally, the main lessons drawn from this joint validation task are summarised, along with an evaluation of the current physical modelling relevance and thus an identification of the ASTEC V2.0 validity domain. © 2013 Elsevier B.V.

SAFEST (Severe Accident Facilities for European Safety Targets) is a European project networking the European corium experimental laboratories with the objective to establish coordination activities, enabling the development of a common vision and research roadmaps for the next years, and of the management structure to achieve these goals. In this frame, a European roadmap on corium experimental research has been written to define research challenges to contribute to further reinforcement of Gen II and III NPP safety. It is based on the research priorities determined by SARNET SARP group as well as those from the recently formulated in the NUGENIA Roadmap for severe accidents and the recently published NUGENIA Global Vision report. It also takes into account issues identified in the analysis of the European stress tests and from the interpretation of the Fukushima accident. 19 relevant issues related to corium have been selected during these prioritization efforts. These issues have been compared to a survey of the European corium experimental facilities and corium analysis laboratories. Finally, the coherence between European infrastructures and R&D needs has been assessed and a table linking issues and infrastructures has been derived. It shows a few lacks in EU corium infrastructures, especially in the domains of core late reflooding impact on source term, Reactor Pressure Vessel failure and corium release, Spent Fuel Pool accidents, as well as the need for a large mass (100–500 kg) prototypic corium facility.

Sixteen laboratories have performed electrochemical noise (EN) measurements based on two systems. The first uses a series of dummy cells consisting of a “star” arrangement of resistors in order to validate the EN measurement equipment and determine its baseline noise performance, while the second system, based on a previous round-robin in the literature, examines the corrosion of aluminium in three environments. All participants used the same measurement protocol and the data reporting and analysis were performed with automatic procedures to avoid errors. The measurement instruments used in the various laboratories include commercial general-purpose potentiostats and custom-built EN systems. The measurements on dummy cells have demonstrated that few systems are capable of achieving instrument noise levels comparable to the thermal noise of the resistors, because of its low level. However, it is of greater concern that some of the instruments exhibited significant artefacts in the measured data, mostly because of the absence of anti-aliasing filters in the equipment or because the way it is used. The measurements on the aluminium samples involve a much higher source noise level during pitting corrosion, and most (though not all) instruments were able to make reliable measurements. However, during passivation, the low level of noise could be measured by very few systems. The round-robin testing has clearly shown that improvements are necessary in the choice of EN measurement equipment and settings and in the way to validate EN data measured. The results emphasise the need to validate measurement systems by using dummy cells and the need to check systematically that the noise of the electrochemical cell to be measured is significantly higher than the instrument noise measured with dummy cells of similar impedance., JRC.F.4-Innovative Technologies for Nuclear Reactor Safety

International audience; One outcome of the OECD/NEA ISP-47 activity was the recommendation to elaborate a 'Generic Containment' in order to allow comparing and rating the results obtained by different lumped-parameter models on plant scale. Within the European SARNET2 project (http//www.sar-net.eu), such a Generic Containment nodalisation, based on a German PWR (1300 MWel), was defined. This agreement on the nodalisation allows investigating the remaining differences among the results, especially the 'user-effect', related to the modelling choices, as well as fundamental differences in the underlying model basis in detail. The methodology applied in order to compare the different code predictions consisted of a series of three benchmark steps with increasing complexity as well as a systematic comparison of characteristic variables and observations. This paper summarises the benchmark series, the lessons learned during specifying the steps, comparing and discussing the results and finally gives an outlook on future steps. © 2014 Elsevier Ltd. All rights reserved.

The influence of containment sprays on atmosphere behaviour, a sub-task of the Work Package WP12-2 CAM (Containment Atmosphere Mixing), has been investigated through benchmark exercises based on TOSQAN (IRSN) and MISTRA (CEA) experiments. These tests are being simulated with lumped-parameter (LP) and Computational Fluid Dynamics (CFD) codes. Both atmosphere depressurization and mixing are being studied in two phases: a 'thermalhydraulic part', which deals with depressurization by sprays (TOSQAN 101 and MISTRA MASPn), and a 'dynamic part', dealing with light gas stratification break-up by spray (TOSQAN 113 and MISTRA MARC2b). In the thermalhydraulic part of the benchmark, participants have found the appropriate modelling to obtain good global results in terms of experimental pressure and mean gas temperature, for both TOSQAN and MISTRA tests. It can thus be considered that code users have a good knowledge of their spray modelling parameters. On a local level, for the TOSQAN test, single droplet behaviour is found to be well estimated by some calculations, but the global modelling of multiple droplets, i.e. of the spray, specifically for the spray dilution, is questionable in some CFD calculations. It can lead to some discrepancies localized in the spray region and can thus have a high impact on the global results, since most of the heat and mass transfers occur inside this region. In the MISTRA tests, wall condensation mass flow rates and local temperatures were used for code-experiment comparison and show that improvement of the local modelling, including initial conditions determination, is needed. In this dynamic part, a general result, in both tests, is that calculations do not recover the same kinetics of the mixing. Furthermore, concerning global mixing, LP contributions seem not suitable here. For the TOSQAN benchmark, the one-phase CFD calculations recover partially the phenomena involved during the mixing, whereas the two-phase flow CFD contributions generally recover the phenomena. Moreover, one important result is also that none of the contributions finds the exact amount of helium remaining in the dome above the spray nozzle in the TOSQAN 113. Discrepancies are rather high (above 5%vol of helium). Results are thus encouraging, but the level of validation should be improved. The same kind of conclusions can be drawn for the MISTRA MARC2B tests. As a conclusion of this SARNET spray benchmark, the level of validation obtained here is encouraging for the use of spray modelling for risk analysis. However, some more detailed investigations are needed to improve model parameters and decrease the uncertainty for containment applications as well as to increase the predictability of the phenomena within the containment analyses. Further activities are well encouraged on this topic, such as numerical benchmarks on analytical separate-effect experiments. © 2011 Elsevier B.V. All rights reserved.

The Phebus Fission Product (FP) program studies key phenomena and phenomenology of severe accidents in water-cooled nuclear reactors. In the framework of the Phebus program, five in-pile experiments were performed that cover fuel rod degradation and behavior of FPs released via the coolant circuit into the containment vessel. Analyses of FP behavior were performed using standard stand-alone versions of codes with input data mainly taken from measured boundary conditions. The FPT2 test used 33 GWd/t uranium dioxide fuel enriched to 4.5%, reirradiated in situ for 7 days to a burnup of 130 MWd/t. This test was designed to study low-pressure FP release and transport through a primary cooling system that included a noncondensing steam generator, with release into the containment vessel in steam-poor conditions. This test also investigated how diluted boric acid in the injected steam influenced FP speciation. In the containment vessel, the objective was to study iodine chemistry in an alkaline sump under evaporating conditions. The analytical approach consisted of progressive studies to explore and explain the main disagreements between base-case calculations and experimental results. Regarding releases in slightly degraded fuel zones, the fission gas behavior and characteristics are shown to be satisfactorily reproduced by the calculations. Electron microprobe analyses also validate the mechanisms for Mo and Ba releases, while the Cs mechanism requires further investigation. Concerning the transport of FPs, a strong connection is shown to exist between Cs, I, Mo, and Cd that substantially impacts vapor-phase chemistry in equilibrium and iodine volatility. Most of the cesium released from fuel is shown to rapidly convert into cesium borates and then into cesium molybdates when the molybdenum release becomes significant at the end of the hydrogen production phase. The main predicted iodine vapor species is cesium iodide. A low fraction of gaseous hydrogen iodide is also calculated at low temperature, but this fraction was found to be strongly dependent on the Cd release kinetics. Hydrogen iodide is the main candidate predicted by equilibrium chemistry calculations to explain the persistence at low temperatures of volatile iodine. Nevertheless, potential limitations on chemical kinetics in the primary circuit zones, characterized by a sharp decrease in temperature, could also be an explanation and are currently under investigation. In the containment, gas-phase reactions were found to predominate in governing iodine chemistry. As for the previous Phebus tests, the gaseous iodine fraction measured in the containment early in the test is thought to come from the primary circuit. However, this low gaseous iodine was hardly tractable by the few dedicated samplings mounted in the primary circuit cold leg upstream from the containment entrance. By favoring hydrolysis reactions of volatile iodine species, the alkaline sump is shown to act as an iodine trap despite the evaporating conditions that prevail during the long-term chemistry phase.however, during this latter phase, a persistent, low-level concentration of gaseous iodine was reached in the long term, as during the previous fpt0/1 tests, indicative of a competition in the containment between iodine traps and sources. Aside from the aerosol particles injected by the primary circuit, in situ iodine oxide particles were found to be continuously forming from the decomposition of i2 and ich3 by air radiolysis products. These particles are suspected to be fine, implying that they predominantly deposit by diffusion on all the containment surfaces. Therefore, in the long term, both the persistence of gaseous iodine and the survival of iodine oxide particles are shown to exist in the containment.

Modelling of corium behaviour during a severe accident requires knowledge of the phases present at equilibrium for a given corium composition, temperature and pressure. The thermodynamic database NUCLEA in combination with a Gibbs Energy minimizer is the European reference tool to achieve this goal. This database has been improved thanks to the analysis of bibliographical data and to EU-funded experiments performed within the SARNET network, PLINIUS as well as the ISTC CORPHAD and EVAN projects. To assess the uncertainty range associated with Energy Dispersive X-ray analyses, a round-robin exercise has been launched in which a UO2-containing corium-concrete interaction sample from VULCANO has been analyzed by three European laboratories with satisfactorily small differences. © 2009 Elsevier Ltd. All rights reserved.

The Phebus FP in-reactor integral experiments provided new insights into iodine transport through the primary circuit. Indeed, in these tests transported iodine was often found not associated with caesium as generally postulated up to now. Several iodine species were experimentally shown to have been transported in the hot leg at 700 °C, while a fraction was also suspected to be in a gaseous form in the cold leg at 150 °C. For a better estimate of the iodine source term to the containment, both in terms of speciation and quantity, it becomes thus necessary to reconsider iodine species behaviour along their pathway in the reactor coolant system (RCS). This paper presents the current understanding, mainly based on SOPHAEROS equilibrium chemistry calculations of Phebus FP tests performed within the I-RCS technical circle of the SARNET network of excellence in the EU 6th Framework programme. The suspected connection existing between Cs, Mo, Cd and I chemistry and the strong influence of both their release kinetics and related species thermodynamic properties on the iodine speciation in different environments (reducing/oxidizing) are highlighted. Potential explanations for the predicted iodine volatility and the level of association of I to Cs are also discussed. © 2009 Elsevier B.V. All rights reserved.

International audience; An international double-blind and blind code benchmark was conducted in the frame of the German THAI program. The basis for the benchmark was test TH-27 which was performed as the commissioning test for the extended, two-vessel THAI+ test facility. The test provides code validation data for simulating flow and mixing inside a containment test facility under typical accident conditions which are of practical interest to nuclear reactor safety. The TH-27 benchmark addresses the phenomena of gas mixing, steam condensation and stratification behavior of light gases under typical containment flow conditions. Double-blind, blind and open simulations were performed by thirteen participants using lumped parameter (LP) models (ASTEC, COCOSYS, MELCOR) as well as CFD codes, namely FLUENT, CFX, GASFLOW and GOTHIC. The challenge of the double-blind benchmark was to generate a model and simulate a long lasting transient without previous model calibration based on knowledge gained from earlier tests being performed in the facility. The experimental measurements allow quantifying transport mechanisms and flow conditions between the two vessels as well as inhomogeneities of the gas mixtures of the vessels. Overall, the TH-27 benchmark demonstrated the high prediction quality of LP and CFD codes provided that dedicated nodalization guidelines are considered. On the other hand the benchmark revealed noticeable user influence for LP codes, especially due to different nodalization schemes and partial lack of special treatment regarding plumes or upward directed jets. © 2018 Elsevier Ltd