[en] IRSN uses a two-tier approach for development of codes analysing the course of a hypothetical severe accident (SA) in a Pressurized Water Reactor (PWR): on one hand, the integral code ASTEC, jointly developed by IRSN and GRS, for fast-running and complete analysis of a sequence; on the other hand, detailed codes for best-estimate analysis of some phenomena such as ICARE/CATHARE, MC3D (for steam explosion), CROCO and TONUS. They have been extensively used to support the level 2 Probabilistic Safety Assessment of the 900 MWe PWR and, in general, for the safety analysis of the French PWR. In particular the codes ICARE/CATHARE, CROCO, MEDICIS (module of ASTEC) and TONUS are used to support the safety assessment of the European Pressurized Reactor (EPR). The ICARE/CATHARE code system has been developed for the detailed evaluation of SA consequences in a PWR primary system. It is composed of the coupling of the core degradation IRSN code ICARE2 and of the thermal-hydraulics French code CATHARE2. The CFD code CROCO describes the corium flow in the spreading compartment. Heat transfer to the surrounding atmosphere and to the basemat, leading to the possible formation of an upper and lower crust, basemat ablation and gas sparging through the flow are modelled. CROCO has been validated against a wide experimental basis, including the CORINE, KATS and VULCANO programs. MEDICIS simulates MCCI (Molten-Corium-Concrete-Interaction) using a lumped-parameter approach. Its models are being continuously improved through the interpretation of most MCCI experiments (OECD-CCI, ACE...). The TONUS code has been developed by IRSN in collaboration with CEA for the analysis of the hydrogen risk (both distribution and combustion) in the reactor containment. The analyses carried out to support the EPR safety assessment are based on a CFD formulation. At this purpose a low-Mach number multi-component Navier-Stokes solver is used to analyse the hydrogen distribution. Presence of air, steam and hydrogen is considered as well as turbulence, condensation and heat transfer in the containment walls. Passive auto-catalytic recombiners are also modelled. Hydrogen combustion is afterwards analysed solving the compressible Euler equations coupled with combustion models. Examples of on-going applications of these codes to the EPR safety analysis are presented to illustrate their potentialities. (authors)

[en] The articles address the different technological perspectives for the development of energy storage: evolutions induced by the policy of energy transition, perspectives and challenges to be faced for the storage of renewable energies, differences and complementarity between storage and cut-off, the future and limitations of energy storage with batteries, progress in solar energy storage in a chemical fluid, the development of offshore energy storage by Tesla, and commitment of the Renault-Nissan group in the energy storage sector

[en] In the case of a severe accident with vessel melt-through, the Molten Core Concrete Interaction (MCCI) may take place and become a threat to the integrity of the containment, which is the ultimate barrier between the corium and the environment. During SARNET project (EURATOM 6th Framework programme) 4 1/2 years, experimental programs related to MCCI issues have been pursued on the VULCANO, HECLA, COMETA and ARTEMIS facilities and several benchmark exercises have been conducted, in particular on reactor scale applications. This issue has been considered within the SARNET project (2004-2008) to have a high priority. Within the new project SARNET2 (2009-2013), a significant experimental program on MCCI, coupled with joint interpretation, modelling and code applications, has been proposed in view of understanding these phenomena and towards closing the safety issue. This research program has been designed to ensure complementarity with the ongoing MCCI-2 project of the OECD-NEA. It will address the following issues: effect of the concrete nature on 2D ablation profiles, role of the metallic layer on the MCCI, and efficiency of water cooling to terminate the ablation of concrete. It will also have a specific focus on the transposition of R and D results to the reactor scale. More than 17 European organizations will contribute to these activities, which will include a significant experimental program both in simulant and in prototypic materials. (author)

[en] This reference document proposes a large and detailed review of severe core fusion accidents occurring in nuclear power reactors. It aims at presenting the scientific aspects of these accidents, a review of knowledge and research perspectives on this issue. After having recalled design and operation principles and safety principles for reactors operating in France, and the main studied and envisaged accident scenarios for the management of severe accidents in French PWRs, the authors describe the physical phenomena occurring during a core fusion accident, in the reactor vessel and in the containment building, their sequence and means to mitigate their effects: development of the accident within the reactor vessel, phenomena able to result in an early failure of the containment building, phenomena able to result in a delayed failure with the corium-concrete interaction, corium retention and cooling in and out of the vessel, release of fission products. They address the behaviour of containment buildings during such an accident (sizing situations, mechanical behaviour, bypasses). They review and discuss lessons learned from accidents (Three Mile Island and Chernobyl) and simulation tests (Phebus-PF). A last chapter gives an overview of software and approaches for the numerical simulation of a core fusion accident

[en] For over thirty years, IPSN and subsequently IRSN has played a major international role in the field of nuclear power reactor core melt accidents through the undertaking of important experimental programmes (the most significant being the Phebus- FP programme), the development of validated simulation tools (the ASTEC code that is today the leading European tool for modelling severe accidents), and the coordination of the SARNET (Severe Accident Research Network) international network of excellence. These accidents are described as 'severe accidents' because they can lead to radioactive releases outside the plant concerned, with serious consequences for the general public and for the environment. This book compiles the sum of the knowledge acquired on this subject and summarises the lessons that have been learnt from severe accidents around the world for the prevention and reduction of the consequences of such accidents, without addressing those from the Fukushima accident, where knowledge of events is still evolving. The knowledge accumulated by the Institute on these subjects enabled it to play an active role in informing public authorities, the media and the public when this accident occurred, and continues to do so to this day

[en] Highlights: • Simple scaling rules and approaches for MCCI are outlined and discussed. • A detailed MCCI code is compared with simplified approaches. • The assumption of a quasi-steady-state energy balance is an adequate simplification. • Simplified procedures are a useful code-independent scaling-up method for MCCI. - Abstract: In the course of a severe accident in a light water reactor, the interactions of corium with the concrete structures of the reactor cavity (Molten Corium–Concrete Interactions or MCCI) may have a significant impact on the long-term integrity of the containment. The 2D behaviour of the melt pool contained in the reactor cavity under dry or top flooding conditions is considered as one of the key phenomena. The “scaling” issue is usually resolved by – in a first step – identifying the impact of physical mechanisms on the process and – in a second step – evaluating these mechanisms at scaled conditions regarding time and length. The conditions for the MCCI change with time due to the evolution of the melt’s state defined by e.g., its composition, temperature and solid fraction, and due to the change of cavity contour and the decreasing decay heat. Here, simplified models are investigated with the objective to infer from laboratory-scale experiments how basic and important parameters like the temperature of the melt and the erosion depth evolve with time if transposed to reactor scale. Due to the simplifications in the models under consideration, the MCCI is analysed assuming “ideal” boundary conditions as e.g., an evolution of a cavity contour with time while retaining its geometrical shape (sphere, cylinder, etc.). Based on these idealised assumptions, generic trends for physical parameters like melt temperature, heat flux at the pool boundary surface, concrete fraction in the melt, viscosity, etc. can be deduced. Simple scaling methods are introduced and checked for consistency by comparison calculations with the MCCI MEDICIS module of the ASTEC integral code. Finally they are applied to a scaling problem under ideal and simplified initial and boundary conditions and the resulting generic trends of the physical parameters are evaluated at reactor scale. Such methods are very useful to better understand the MCCI phenomenology although more detailed MCCI codes are indispensable to simulate more complex accident sequences or to take into account complex boundary conditions

[en] Activities carried out over the last three decades in relation to core-concrete interactions and melt coolability, as well as related containment failure modes, have significantly increased the level of understanding in this area. In a severe accident with little or no cooling of the reactor core, the residual decay heat in the fuel can cause the core materials to melt. One of the challenges in such cases is to determine the consequences of molten core materials causing a failure of the reactor pressure vessel. Molten corium will interact, for example, with structural concrete below the vessel. The reaction between corium and concrete, commonly referred to as MCCI (molten core concrete interaction), can be extensive and can release combustible gases. The cooling behaviour of ex-vessel melts through sprays or flooding is also complex. This report summarises the current state of the art on MCCI and melt coolability, and thus should be useful to specialists seeking to predict the consequences of severe accidents, to model developers for severe-accident computer codes and to designers of mitigation measures

[en] For over thirty years, IPSN and subsequently IRSN has played a major international role in the field of nuclear power reactor core melt accidents through the undertaking of important experimental programmes (the most significant being the Phebus-FP programme), the development of validated simulation tools (the ASTEC code that is today the leading European tool for modelling severe accidents), and the coordination of the SARNET (Severe Accident Research Network) international network of excellence. These accidents are described as 'severe accidents' because they can lead to radioactive releases outside the plant concerned, with serious consequences for the general public and for the environment. This book compiles the sum of the knowledge acquired on this subject and summarises the lessons that have been learnt from severe accidents around the world for the prevention and reduction of the consequences of such accidents, without addressing those from the Fukushima accident, where knowledge of events is still evolving. The knowledge accumulated by the Institute on these subjects enabled it to play an active role in informing public authorities, the media and the public when this accident occurred, and continues to do so to this day. Following the introduction, which describes the structure of this book and highlights the objectives of R and D on core melt accidents, this book briefly presents the design and operating principles (Chapter 2) and safety principles (Chapter 3) of the reactors currently in operation in France, as well as the main accident scenarios envisaged and studied (Chapter 4). The objective of these chapters is not to provide exhaustive information on these subjects (the reader should refer to the general reference documents listed in the corresponding chapters), but instead to provide the information needed in order to understand, firstly, the general approach adopted in France for preventing and mitigating the consequences of core melt accidents and, secondly, the physical phenomena, studies and analyses described in Chapters 5 to 8. Chapter 5 is devoted to describing the physical phenomena liable to occur during a core melt accident, in the reactor vessel and the reactor containment. It also presents the sequence of events and the methods for mitigating their impact. For each of the subjects covered, a summary of the physical phenomena involved is followed by a description of the past, present and planned experiments designed to study these phenomena, along with their modelling, the validation of which is based on the test results. The chapter then describes the computer codes that couple all of the models and provide the best current state of knowledge of the phenomena. Lastly, this knowledge is reviewed while taking into account the gaps and uncertainties, and the outlook for the future is presented, notably regarding experimental programmes and the development of modelling and numerical simulation tools. Chapter 6 focuses on the behaviour of the containment enclosures during a core melt accident. After summarising the potential leakage paths of radioactive substances through the different containments in the case of the accidents chosen in the design phase, it presents the studies of the mechanical behaviour of the different containments under the loadings that can result from the hazards linked with the phenomena described in Chapter 5. Chapter 6 also discusses the risks of containment building bypass in a core melt accident situation. Chapter 7 presents the lessons learned regarding the phenomenology of core melt accidents and the improvement of nuclear reactor safety. Lastly, Chapter 8 presents a review of development and validation efforts regarding the main computer codes dealing with 'severe accidents', which draw on and build upon the knowledge mainly acquired through the research programmes: ASTEC (IRSN and GRS), MAAP-4 (FAI (US)) and used by EDF and by utilities in many other countries, and MELCOR (SNL (US)) for the US Nuclear Regulatory Commission (US NRC)