[en] Following the events at Fukushima, the Institute for Radiological Protection and Nuclear Safety (IRSN) has been strongly involved in a series of reviews related to the robustness of French nuclear power plants in case of 'rare and severe' external hazards. These reviews included in particular the 'stress tests' performed in 2011 as required by the European Commission. Those reviews, and the proposal made by EDF to reinforce NPPs robustness in such situation, led to the introduction of the concept of a hardened safety core (HSC) to avoid massive releases and prolonged effects in the environment in case of rare and severe natural hazards. This concept will be explained in the paper and the new specific electrical equipment as well as the interfaces with the existing electrical distribution required to implement this HSC will be explained. As the detailed design, manufacturing and installation of the HSC in all NPP sites will take several years, temporary measures have been adopted. This paper will also present the electrical sources and the distribution related to those temporary measures. The specific situation of the new built EPR reactor in Flamanville is also addressed. Lastly, in complement to the above on-site design provisions, a Nuclear Rapid Response Force has been set up by EDF to bring off-site support to French NPPs in case of emergency. The paper will describe the type of electrical equipment to be delivered and the principle for distributing the electrical power to the required loads. (authors)

[en] The total loss of the ultimate heat sink is an initiating event which, even it is mainly of external origin, has been considered in the frame of internal events Level 1 PSA by IRSN. The on-going actions on the development of external hazards PSA and the recent incident of loss of the heat sink induced by the ingress of vegetable matter that occurred in France in 2009 have pointed out the need to improve the modeling of the loss of the heat sink initiating event and sequences to better take into account the fact that this loss may be induced by external hazards and thus affect all the site units. The paper presents the historical steps of the modeling of the total loss of the heat sink, the safety stakes of this modeling, the main assumptions used by IRSN in the associated PSA for the 900 MWe reactors and the results obtained. The total loss of the heat sink was not initially addressed in the safety demonstration of French NPPs. On the basis of the insights of the first probabilistic assessments performed in the 80's, the risks associated to this 'multiple failure situation' turned out to be very significant and design and organisational improvements were implemented on the plants. Reviews of the characterization of external hazards and of their consequences on the installations and French operating feedback have revealed that extreme hazards may induce a total loss of the heat sink. Moreover, the accident that occurred at Fukushima in 2011 has pointed out the risk of such a loss of long duration at all site units in case of extreme hazards. In this context, it seems relevant to further improve the modelling of the total loss of the heat sink by considering the external hazards that may cause this loss. In a first step, IRSN has improved the assumptions and data used in the loss of the heat sink PSA model, in particular by considering that such a loss may affect all the site units. The next challenge will be the deeper analysis of the impact of external hazards on the equipment necessary to cope with the loss of the ultimate heat sink. (authors)

[en] In the traditional Level 1 PSA, the long term of the accident sequences is usually taken into account in a simplified manner. For example, some of the mitigations which are needed at long term are taken into account in the PSA, but the analysis and the associated failures probabilities quantification are estimated based on generic assessments. In the context of the extension of PSA scope to include the external hazards, in France, both operator (EDF) and IRSN work for the improvement of methods to better take into account in the PSA the long term of accident sequences induced by initiators which affect the whole site containing several nuclear installations (reactors, fuel pools, ...). This is an essential prerequisite for the development of external hazards PSA. It has to be noted that in the French PSA, even before Fukushima, this type of accident sequences was already taken into account, many insight being used, as complementary information, to enhance the safety level of the plants. The recent French and international operating experience is an opportunity for tuning the actual PSA methods for long term accident sequences modeling. The paper presents the main results of the ongoing efforts in this area. (author)

[en] After an overview by the ASN of complementary safety assessments and an assessment of 'post-Fukushima' inspections of basic nuclear installations, the contributions (Power Point presentations) of this seminar proposed: the opinion of the Gravelines CLI (local information commission) on the Gravelines complementary safety assessment report, an analysis and discussion by the GSIEN on reports of complementary assessment of safety of nuclear installations with respect to the Fukushima accident, an analysis by the IRSN of complementary safety assessments performed by operators, the IRSN approach to analyze complementary safety assessments, reports on installation conditions, external flooding and seismic hazard, 'meltdown prevention' aspects in the management of accidental situations in EDF reactors

[en] After a synthetic report of the meeting, this document contains Power Point presentations proposed by the different contributors. These presentations proposed: an overview on additional safety assessments (ECS) and an assessment of 'post-Fukushima' inspections performed in basic nuclear installations; the CLI's opinion on the Gravelines ECS report; an analysis and a discussion of ECS reports of nuclear installations in the perspective of the Fukushima accident; the IRSN analysis of ECS as they are performed by operators; a presentation of the IRSN analysis approach to ECS; contributions of different post-Fukushima ECS permanent groups within the IRSN (these work groups address installation condition, external flooding, and seismic risk); a presentation of the 'fusion prevention' aspects of the management of accidental situations in EDF reactors

[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)