[en] This thesis characterized a mixer hydrodynamic, using the Couette-Taylor properties in the case of one-phase, two-phase (air-liquid) and three-phase (air-liquid-liquid). An ideal configuration has been defined. This study brings a contribution to the fuels processing processes.

[en] In-vessel retention (IVR) of molten corium through external cooling of the reactor pressure vessel is one possible means of severe accident mitigation for a class of nuclear power plants. The aim is to successfully terminate the progression of a core melt within the reactor vessel. The probability of success depends on the efficacy of the cooling strategy; hence one of the key aspects of an IVR demonstration relates to the heat removal capability through the vessel wall by convection and boiling in the external water flow. This is only possible if the in-vessel thermal loading is lower than the local critical heat flux expected along the outer wall of the vessel, which is in turn highly dependent on the flow characteristics between the vessel and the insulator. The NEPTUNE-CFD multiphase flow solver is used to obtain a better understanding at local scale of the thermal hydraulics involved in this situation. The validation of the NEPTUNE-CFD code on the ULPU-V facility experiments carried out at the University of California Santa Barbara is presented as a first attempt of using CFD codes at EDF to address such an issue. Two types of computation are performed. On the one hand, a steady state algorithm is used to compute natural circulation flow rates and differential pressures and, on the other, a transient algorithm computation reveals the oscillatory nature of the pressure data recorded in the ULPU facility. Several dominant frequencies are highlighted. In both cases, the CFD simulations reproduce reasonably well the experimental data for these quantities.

[en] Within the framework of severe reactor accident studies, we present experimental and numerical parametric studies on debris bed coolability. Data are provided by the SILFIDE multidimensional experimental facility at Electricite de France. The bed is composed of inductively heated steel sphere beads (diameters ranging from 2 to 7.18 mm) contained in a 50- x 60- x 10-cm vessel. Numerical computations are obtained with MC3D REPO developed by Commissariat a l'Energie Atomique.Because of heterogeneous power distribution within the bed, two definitions (mean and local) for the critical heat flux (CHF) are proposed. Even in the first case, the CHF was higher than the Lipinsky one-dimensional flux. As the power is being increased, temperature plateaus above saturation temperature are observed. An analysis is proposed, based on possible different hydrodynamic flow configurations occurring in postdryout regimes. In some experiments, some spheres were superficially molten and stacked together, but globally, the bed was still coolable.The influence of operational parameters such as bottom coolant injection, height of the water, fluidization of upper particles, and subcooled liquid injection on dryout phenomena and CHF values are also described.The MC3D-REPO calculations assuming a thermal equilibrium between the three phases gives results in accordance with experimental data

[en] Highlights: • Low-carbon steel samples pre-oxidized in high-temperature, humid-air environment. • Reproduced prototypical surface oxides present on reactor pressure vessel surface. • Flow CHF measured at thermal-hydraulic conditions and chemistry relevant to IVR. • Large CHF enhancement observed. • Empirical correlation fitting the CHF data presented. - Abstract: In severe accident mitigation approaches that aim to achieve In-Vessel Retention (IVR) the decay heat is removed from the corium by conduction through the Reactor Pressure Vessel (RPV) wall, and by flow boiling on the outer surface of the RPV. The boiling Critical Heat Flux (CHF) limit must not be exceeded to prevent RPV failure. Previous studies for prediction of CHF in IVR were predominantly based on data for stainless-steel heaters and de-ionized (DI) water coolant. However, the RPV is made of low-carbon steel, and its surface has an oxide layer that results from pre-service heat treatment as well as oxidation during service; this oxide layer renders the surface much more hydrophilic and rough with respect to an un-oxidized stainless-steel surface, which can have a significant influence on boiling heat transfer. In this study, test heaters were fabricated from low-carbon steel (grade 18MnD5), pre-oxidized in a controlled, high-temperature, humid-air environment, reproducing the prototypical surface oxides present on the outer surface of the RPV. The heaters were then tested in a flow boiling loop using the IVR water chemistry, i.e., DI water with addition of boric acid and sodium tetra-borate. CHF was measured in the range of pressures (100–440 kPa), mass fluxes (180–2450 kg/m² s), inclination angles (30–90°) and equilibrium qualities (from −0.020 to +0.034) encompassing the IVR conditions. Up to 70% enhancement in CHF values was observed for pre-oxidized, low-carbon steel heaters in comparison to the stainless-steel control heaters. The effect of water chemistry on the CHF was found to be marginal. An empirical correlation fitting the CHF data for pre-oxidized, low-carbon steel surfaces with IVR water chemistry is also presented.

[en] Benchmarking work was recently performed for the issue of molten corium concrete interaction (MCCI). A synthesis is given here. It concerns first the 2D CCI-2 test with a homogeneous pool and a limestone concrete, which was used for a blind benchmark. Secondly, the COMET-L2 and COMET-L3 2D experiments in a stratified configuration were used as a post-test (L2) and a blind-test (L3) benchmark. More details are given here for the recent benchmark considering a matrix of four reactor cases, with both a homogeneous and a stratified configuration, and with both a limestone and a siliceous concrete. A short overview is given on the different models used in the codes, and the consistency between the benchmark actions on experiments and reactor situations is discussed. Finally, the major uncertainties concerning MCCI are also pointed out. (authors)

[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)