Combustible nucléaire

Résumé

vignette|Modèle de l'atome. Le combustible nucléaire est le produit qui, contenant des isotopes fissiles (uranium, plutonium…), fournit l'énergie dans le cœur d'un réacteur nucléaire en entretenant la réaction en chaîne de fission nucléaire. Les termes « combustible » et « combustion » sont utilisés par analogie à la chaleur dégagée par une matière en feu, mais sont inappropriés pour caractériser tant le produit que son action. En effet, la combustion est une réaction d'oxydoréduction (échange d'électrons) tandis que la « combustion » des matières radioactives provient de réactions nucléaires (fission de noyaux atomiques). Les matières fissiles sont utilisées pour la propulsion nucléaire de navires militaires (en particulier de porte-avions et de sous-marins nucléaires), ainsi que comme combustible dans les centrales nucléaires. Un réacteur à eau pressurisée de comporte environ de combustible renouvelé périodiquement, par parties. Le combustible UOX (Uranium Oxide) est constitué

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Zirconium alloys used in the nuclear industry are exposed to extreme conditions undergoing high levels of irradiation damage and corrosion. Zircaloy-2 is used as nuclear fuel cladding in boiling water reactors and for the encapsulation of the spallation target in the neutron source SINQ at the Paul Scherrer Institute. As a result of oxidation, hydrogen is produced and partially taken up into the zirconium matrix. As the hydrogen reaches the solubility limit, it precipitates in the form of a zirconium hydride. A number of deleterious effects of hydrides have been extensively studied, including the embrittlement of the material and a unique cracking phenomenon, namely delayed hydride cracking (DHC). DHC occurs as hydrogen in solid solution diffuses towards locations of higher tensile stress, where it subsequently precipitates as the local solvus limit is reached. Once the hydride grows to a critical size, it fractures, which can continue to propagate in repeated steps by the same mechanism.Several studies have been performed in order to understand properties of DHC based on parameters like, alloy type, cracking temperature, hydrogen content, and cracking direction. The most important mechanical properties include the fracture toughness and cracking velocity, which governs the material failure conditions. Crystallographic properties have been investigated in order to understand how complex hydride precipitation develops within the context of DHC conditions. In this project, radial DHC is explored in irradiated and unirradiated thin-walled Zircaloy-2 nuclear fuel cladding with and without an inner liner, in addition to spallation target rod material. A unique three-point bending test, applicable also for irradiated samples in the hotlab, was developed to induce a consistent radially propagating outside-in crack. Results have provided insight into the effects of the inner liner on cracking velocity, showing correlations with the hydrogen depletion. The combined DHC and creep mechanisms, which cannot be decoupled, is coined as âcreep-delayed hydride crackingâ. As the initiating conditions for DHC are of great technical importance, the threshold stress intensity factor has been investigated, under ideal DHC conditions, where a minimum value occurs around 6 MPaâm. Quantitative analysis of the hydrogen distribution around the tip of an arrested DHC crack tip has been performed with neutron radiography at the SINQ. Trends show that the hydrogen concentration around the crack tip increases with temperature resulting likely from the larger hydrogen diffusion, and that the liner reduces the hydrogen available for DHC. Crystallographic analysis has provided insight on the hydride phases responsible for DHC with synchrotron micro X-ray diffraction, performed at the Swiss Light Source at PSI. The results directly suggest that the Î³-hydride is a stable phase precipitated at low temperatures. Testing of irradiated material provided insight into the crack initiation of DHC as a result of the strongly irregular outer surface generated from the oxidation. The propagation pattern is highly irregular compared to the unirradiated material due to the embrittled nature of the cladding from irradiation damage.

EPFL2022

Chargement

Study of Microfluidic Mixing and Droplet Generation for 3D Printing of Nuclear Fuels

Internal gelation and 3D printing processes were proposed to combine in a process that is promising to produce nuclear fuels with simplified production route and enhanced in-core performance [1]. The process relies on in-situ mixing of feed solutions that bear actinides and uranium nitrates with the methenamine and urea mixture. These feed solutions should be thoroughly mixed and kept at low temperature up to an inkjet printing head, where an inkjet printing process is used to deposit controlled volumes of the mixed solution on the pellet under-fabrication. Microwave and/or laser-heating are then used to trigger the gelation reaction and to solidify these printed dots and later to drive out the excess water and pre-sinter the gelled oxides. The current study is an initial approach to optimize the upfront parameters of the 3D printing process. The feed solutions need be mixed in a passive microfluidic mixer, where the first optimization parameter is the minimum mixing time in the microfluidic mixer that will yield an acceptable mixing level. To do so, CFD modelling using OpenFOAM for different designs of microfluidic mixers was conducted to select the most efficient conception. To validate the CFD results, simulation of microfluidic mixers similar to some literature cases were performed and compared. The uncertainty of selected parameters in the CFD models was used to evaluate a conservative retention time and length. Secondly, the next printing step was modeled with OpenFOAM, which is the droplet generation process. The objective from these models was to correlate the feed parameters like the flow rate and the usage of an acoustic mechanical vibrator to the printed droplets size, generation rate, and the required stand-off distance of the printing head. To perform these CFD simulations with the right solution parameters, rheology and diffusivity measurements were conducted. A proposed empirical approach was investigated for the evaluation of the diffusivity of multi-species solutions. The resultant parameters from the current study are the mixing times and lengths for given microfluidic mixers that are industrially feasible, also, the rate and volume at which the droplets of the broth mixture are generated. These results can be used as guidelines for designing the inkjet printing head and to highlight areas where further investigations are required for an efficient production process.

EPFL2017

Chargement

Conceptual Design of a Sustainable Waste Burning Molten Salt Reactor

Boris Aviv Hombourger

For an energy source to qualify as sustainable, it must maximize the efficiency with which it uses natural resources while minimizing the amounts of waste it produces. Currently deployed nuclear power plants however use a very limited amount of the energy contained in natural nuclear fuels such as Uranium and Thorium and produce long-lived nuclear waste that must be dealt with. Molten Salt Reactors (MSRs) are one family of advanced reactors designs with the potential for safe and sustainable energy production due to their use of a liquid molten salt fuel. This thesis aimed at reviewing MSR designs that have been proposed in the past, investigating alternative design possibilities and proposing a concept capable of disposing of existing waste in a sustainable manner. One possibility to do so is to design a reactor to run on an isobreeding closed fuel cycle while using existing actinide waste as a start-up inventory, which is the route investigated in this thesis. For this purpose, the EQL0D fuel cycle procedure was developed using the MATLAB environment and the Serpent 2 Monte Carlo neutron transport code to simulate the start-up, transition and equilibrium of MSRs. The procedure was then applied to investigate the performance at equilibrium of several candidate fuel salts and moderators in an infinite lattice and in a closed Uranium-Plutonium and Thorium-Uranium fuel cycle. The results showed that only an isobreeding closed Uranium-Thorium fuel cycle can be achieved in moderated MSRs, while both isobreeding Uranium-Plutonium and Thorium-Uranium closed fuel cycles can be sustained in fast-spectrum MSRs, which additionally show a substantial reactivity margin that enable them to dispose of poorly fissile nuclides. The transition behavior of selected fluoride- and chloride-fueled reactors was then investigated. The fluoride-fueled cores all feature a closed Thorium-Uranium fuel cycle and included two historical graphite-moderated designs (the single- and two-fluid Molten Salt Breeder Reactors) and a modern fast-spectrum design (Molten Salt Fast Reactor). The chloride-fueled cores were proposed based on the results obtained in the infinite lattice study. The results obtained indicated that the transuranic-burning capabilities of fluoride-fueled reactors are substantially limited by the low solubility of transuranic trifluorides in fluoride salt mixtures, while no such limitation was found in chloride salts. On the other hand, chloride salts necessitate substantially larger cores and inventories. Finally, the possibility of operating MSRs on two types of open cycles without fuel processing, breed-and-burn and once-through, was investigated to alleviate concerns related to the uncertainties related to the technical feasibility of molten salt fuel processing in closed fuel cycles. The results obtained show that it is possible to operate chloride-fueled MSRs in a sustainable breed-and-burn fuel cycle and also on an open cycle with appreciable burn-ups achieved.

EPFL2018

Chargement