[en] Models of crevice corrosion can be divided into two categories: the first one is aimed to define the time necessary to reach a Critical Crevice Solution susceptible to initiate a stable crevice propagation whereas the second one is focused on the chemical composition and potential in the crevice during its steady propagation. In this second category the geometry of the crevice is kept constant which is a very rough approximation since a real crevice never reaches a steady state mainly because of its shape evolution. Such an approach necessitates the determination of the most important input parameters (external solution composition, applied potential, shape of the crevice, etc.) in the stabilization of a crevice providing a stability criterion is defined, taking into account the occurrence of precipitation or of gas evolution. The objective of this study was to determine under which conditions of pH and potential a crevice was susceptible to re-passivate. For doing this we used commercial code, since existing ones are mostly home-made, keeping in mind that it had to be as a modular as possible. This code was developed using the Chemical Engineering Module of FEMLAB, which is a MATLAB-based tool for finite element methods. In a first part of this study the ability of this software to be used for crevice corrosion on iron will be presented. As function of the environment (bulk composition and applied potential), calculations were performed in order to determine the occurrence of solid precipitation like FeCl₂ and Fe(OH)₂ or H₂ gas bubbles generation inside the occluded cavity. (authors)

[en] Because of the synergetic actions of a corrosive medium and mechanical loading which occur in stress corrosion cracking, modelling this kind of damage on actual industrial structures can be very challenging. However, when studying one single couple material/medium, it is possible to only consider the mechanical aspect of damaging. Establishing this sort of purely mechanical simulation implies to have extensive information on the behavior of the propagating cracks. Several types of tests are commonly used to gather experimental data on crack propagation, among which are tests conducted on fracture mechanics sample (CT, WOL, etc.) and slow strain rate tests (SSRT) on smooth tensile specimens. The first category of tests is mainly used to obtain the propagation rate of long cracks and the so-called K_ISCC, the mode I stress intensity factor below which no propagation can be detected for constant displacement tests. The SSRT are preferred to study initiation, mechanisms and short cracks propagation (below K_ISCC). However, as the analyses of the cracking of these tensile tests are essentially statistical, it is somewhat difficult to separate the behavior of a single short crack from mechanisms resulting from cracks interaction. For instance, for alloy 600 in primary water, the transition of crack growth rate from a slow to a fast regime for cracks attaining the K_ISCC can be interpreted as a mechanism existing for the single crack or as a result of a natural selection within the cracks population. In order to obtain experimental data on the propagation of one single short crack, we have developed SSRT on micro-notched tensile specimens. This new kind sample allows the initiation and the propagation of one single crack at notch's tip. The experiments were conducted for an austenitic stainless steel (316L) in a model chloride medium (30% MgCl₂ at 117 deg C). The use of interrupted tests permits us to obtain crack propagation rate in function of the global loading. Modelling these tests, as well as WOL tests, with the finite elements method gives us access to the local mechanical parameters, as strain or strain rate, needed in global stress corrosion cracking simulations. Besides, the predominance of crack branching in SSRT compared to WOL tests has been investigated. The use of EBSD (electron back scattering diffraction) showed the roles of twin and grain boundaries in branching, as well as the confined plasticity on the crack path, revealed by creation of sub-grains boundaries. (authors)

[en] This book first explains the choice of sodium-cooled reactors by outlining the reasons of the choice of fast neutron reactors (fast neutrons instead of thermal neutrons, recycling opportunity for plutonium, full use of natural uranium, nuclear waste optimization, flexibility of fast neutron reactors in nuclear material management, fast neutron reactors as complements of water-cooled reactors), and by outlining the reasons for the choice of sodium as heat-transfer material. Physical, chemical, and neutron properties of sodium are presented. The second part of the book first presents the main design principles for sodium-cooled fast neutron reactors and their core. The third part proposes an historical overview and an assessment of previously operated sodium-cooled fast neutron reactors (French reactors from Rapsodie to Superphenix, other reactors in the world), and an assessment of the main incidents which occurred in these reactors. It also reports the experience and lessons learned from the dismantling of various sodium-cooled fast breeder reactors in the world. The next chapter addresses safety issues (technical and safety aspects related to the use of sodium) and environmental issues (dosimetry, gaseous and liquid releases, solid wastes, and cooling water). Then, various technological aspects of these reactors are addressed: the energy conversion system, main components, sodium chemistry, sodium-related technology, advances in in-service inspection, materials used in reactors and their behaviour, and fuel system. The next chapter addresses the fuel cycle in these reactors: its integrated specific character, report of the French experience in fast neutron reactor fuel processing, description of the transmutation of minor actinides in these reactors. The last chapter proposes an overview of reactors currently projected or under construction in the world, presents the Astrid project, and gives an assessment of the economy of these reactors. A glossary and an index are provided

[en] The control of the corrosion phenomenon is of prime importance for the nuclear industry. The efficiency and the safety of facilities can be affected by this phenomenon. The nuclear industry has to face corrosion for a large variety of materials submitted to various environments. Metallic corrosion operates in the hot and aqueous environment of water reactors which represent the most common reactor type in the world. Progresses made in the control of the corrosion of the different components of these reactors allow to improve their safety. Corrosion is present in the facilities of the back-end of the fuel cycle as well (corrosion in acid environment in fuel reprocessing plants, corrosion of waste containers in disposal and storage facilities, etc). The future nuclear systems will widen even more the range of materials to be studied and the situations in which they will be placed (corrosion by liquid metals or by helium impurities). Very often, corrosion looks like a patchwork of particular cases in its description. The encountered corrosion problems and their study are presented in this book according to chapters representing the main sectors of the nuclear industry and classified with respect to their phenomenology. This monograph illustrates the researches in progress and presents some results of particular importance obtained recently. Content: 1 - Introduction: The context, stakes, and goals; What is corrosion? A few definitions; A complex science; Corrosion in the nuclear field; 2 - Corrosion in Water-Cooled Reactors - Phenomenology, Mechanisms, and Remedies: Uniform Corrosion (Uniform corrosion mechanisms, Uniform corrosion of fuel clads, In situ measurements of general corrosion rate through electrochemical techniques, Uniform corrosion of nickel alloys: characterization of the passive layer and growth mechanisms, The PACTOLE code, an integrating tool, Water chemistry influence on corrosion and contamination, Radiolysis impact on uniform corrosion); Stress Corrosion Cracking (overview, Test tools for investigating stress corrosion cracking, Experimental techniques, Internal corrosion of Zircaloy clads: iodine effect, Stress corrosion cracking of nickel base alloys: hydrogen influence, Stress corrosion cracking of stainless steels); Wear Corrosion (A coupled phenomenon, A research work related to extended lifetime of the French nuclear power reactor fleet); 3 - Corrosion in the Nuclear Reactor Systems of the Future: Corrosion in Gas-Cooled Reactors (Corrosion by helium impurities, Silicon carbide resistance to oxidation, Corrosion of graphite and carbon-carbon composites); Materials Corrosion in Liquid Metal-Cooled Reactors (Corrosion in sodium-cooled fast reactors, Corrosion in lead (alloy)-cooled fast reactors); Corrosion in Molten Salt Reactors (Corrosion of nickel alloys (Hastelloy-N) by molten fluorides, Mass transfer in non-isothermal fluoride systems, Tellurium-induced embrittlement, Electrochemical study of pure metal corrosion in molten fluorides); 4 - Materials Corrosion and Alteration at the Back-End of Fuel Cycle: Corrosion in a Concentrated Nitric Environment (Materials behavior in a nitric environment, Autocatalytic mechanism of nitric acid reduction); Corrosion in an Aqueous Unsaturated Environment (Metal corrosion in an unsaturated environment - The specific case of waste package storage, Bitumen alteration, Reinforced concrete behavior and reinforcement corrosion, Concrete behavior in a severe thermal environment); Corrosion in an Aqueous Saturated Environment (Metal corrosion in a clay environment: the specific case of disposal package containers - What about the liability of occurrence of a gas release?, Long-term behavior of glasses and alteration by water: the specific case of nuclear waste disposal, Ceramic alteration, 'Underwater' durability of concretes, Transformation of engineered barrier clays); Materials Biodegradation (Microorganisms and nuclear waste, Biodegradation of embedding materials: bitumen, Biodegradation of embedding materials: hydraulic binders, Biocorrosion of packaging materials: steels); 5 - Conclusion; Glossary - Index

[en] Recent progress in the field of high-temperature materials and components has put these types of reactors again in the forefront of research and development. After the work already initiated on Magnox reactors, gas-cooled reactors have been considerably modernized and could eventually challenge the current supremacy of PWRs, at least in certain areas or specific applications. Industrial-scale implementations of slow-neutron versions are foreseeable in the medium term. Fast-neutron versions (more prospective) offer additional possibilities for energy-efficient use of natural uranium resources by using a fuel cycle that minimises final waste production and proliferation risks. This monograph covers the current research on these types of reactors, describing the research challenges involved, the recent results achieved at the CEA and the obstacles that remain to be overcome

[en] Fuel is one of the essential components in a nuclear reactor. It is within that fuel that nuclear reactions take place, i.e. fission of heavy atoms, uranium and plutonium. Fuel is at the core of the reactor, but equally at the core of the nuclear system as a whole. Fuel design and properties influence reactor behaviour, performance, and safety. Even though it only accounts for a small part of the cost per kilowatt-hour of power provided by current nuclear power plants, good utilization of fuel is a major economic issue. Major advances have yet to be achieved, to ensure longer in-reactor dwell-time, thus enabling fuel to yield more energy, and improve ruggedness. Aside from economics, and safety, such strategic issues as use of plutonium, conservation of resources, and nuclear waste management have to be addressed, and true technological challenges arise. This monograph surveys current knowledge regarding in-reactor behaviour, operating limits, and avenues for R and D. It also provides illustrations of ongoing research work, setting out a few noteworthy results recently achieved

[en] Published in English and in French, this large report first proposes an overview of the use and history of research nuclear reactors. It discusses their definition, and presents the various types of research reactors which can be either related to nuclear power (critical mock-ups, material test reactors, safety test reactors, training reactors, prototypes), or to research (basic research, industry, health), or to specific particle physics phenomena (neutron diffraction, isotope production, neutron activation, neutron radiography, semiconductor doping). It reports the history of the French research reactors by distinguishing the first atomic pile (ZOE), and the activities and achievements during the fifties, the sixties and the seventies. It also addresses the development of instrumentation for research reactors (neutron, thermal, mechanical and fission gas release measurements). The other parts of the report concern the validation of neutronics calculations for different reactors (the EOLE water critical mock-up, the MASURCA air critical mock-up dedicated to fast neutron reactor study, the MINERVE water critical mock-up, the CALIBAN pulsed research reactor), the testing of materials under irradiation (OSIRIS reactor, laboratories associated with research reactors, the Jules Horowitz reactor and its experimental programs and related devices, irradiation of materials with ion beams), the investigation of accident situations (on the CABRI, Phebus, Silene and Jules Horowitz reactors). The last part proposes a worldwide overview of research reactors

[en] This collective publication proposes presentations of scientific approaches implemented to model and simulate the behaviour of materials submitted to irradiation, of associated experimental methods, and of some recent important results. After an introduction presenting the various materials used in different types of nuclear reactors (PWR, etc.), the effects of irradiation at the macroscopic or at the atomic scale, and the multi-scale (time and space) approach to the modelling of these materials, a chapter proposes an overview of modelling tools: multi-scale approach, electronic calculations for condensed matter, inter-atomic potentials, molecular dynamics simulation, thermodynamic and medium force potentials, phase diagrams, simulation of primary damages in reactor materials, kinetic models, dislocation dynamics, production of microstructures for simulation, crystalline visco-plasticity, homogenization methods in continuum mechanics, local approach and probabilistic approach in material fracture. The next part presents tools for experimental validation: tools for microscopic characterization or for mechanical characterization, experimental reactors and tests in atomic pile, tools for irradiation by charged particles. The next chapters presents different examples of thermodynamic and kinetic modelling in the case of various alloys (zirconium alloys, iron-chromium alloys, silicon carbide, austenitic alloys), of plasticity and failure modelling

[en] This collective publication proposes presentations of scientific approaches implemented to model and simulate the behaviour of materials submitted to irradiation, of associated experimental methods, and of some recent important results. After an introduction presenting the various materials used in different types of nuclear reactors (PWR, etc.), the effects of irradiation at the macroscopic or at the atomic scale, and the multi-scale (time and space) approach to the modelling of these materials, a chapter proposes an overview of modelling tools: multi-scale approach, electronic calculations for condensed matter, inter-atomic potentials, molecular dynamics simulation, thermodynamic and medium force potentials, phase diagrams, simulation of primary damages in reactor materials, kinetic models, dislocation dynamics, production of microstructures for simulation, crystalline visco-plasticity, homogenization methods in continuum mechanics, local approach and probabilistic approach in material fracture. The next part presents tools for experimental validation: tools for microscopic characterization or for mechanical characterization, experimental reactors and tests in atomic pile, tools for irradiation by charged particles. The next chapters presents different examples of thermodynamic and kinetic modelling in the case of various alloys (zirconium alloys, iron-chromium alloys, silicon carbide, austenitic alloys), of plasticity and failure modelling

[en] Subsequent to its in-reactor dwell time, spent fuel still contains large amounts of materials that are recoverable, for value-added energy purposes (uranium, plutonium), together with fission products, and minor actinides, making up the residues from nuclear reactions. The treatment and recycling of spent nuclear fuel, as implemented in France, entail that such materials be chemically partitioned. The development of the process involved, and its deployment on an industrial scale stand as a high achievement of French science, and technology. Treatment and recycling allow both a satisfactory management of nuclear waste to be implemented, and substantial savings, in terms of fissile material. Bolstered of late as it has been, due to spectacularly skyrocketing uranium prices, this strategy is bound to become indispensable, with the advent of the next generation of fast reactors. This Monograph surveys the chemical process used for spent fuel treatment, and its variants, both current, and future. It outlines currently ongoing investigations, setting out the challenges involved, and recent results obtained by CEA