Énergie nucléaire | EPFL Graph Search

Selon le contexte d'usage, le terme d’énergie nucléaire recouvre plusieurs acceptions, toutes liées à la physique et aux réactions de noyaux atomiques.

Dans le langage courant, l’énergie nucléaire correspond aux usages civils et militaires de l’énergie libérée lors des réactions de fission nucléaire ou de fusion nucléaire de noyaux atomiques au sein d'un réacteur nucléaire ou lors d'une explosion atomique.
Dans le domaine des sciences de la Terre et de l'Univers, l'énergie nucléaire est l’énergie libérée par les réactions de fusion nucléaire au sein des étoiles ainsi que par la radioactivité naturelle, la principale source d'énergie du volcanisme de la Terre.
En physique des particules, l’énergie nucléaire est l’énergie associée à la force de cohésion des nucléons (protons et neutrons), la force nucléaire forte au sein du noyau des atomes. Les transformations du noyau libérant cette énergie sont les réactions nucléaires. La force nucléaire faible régit les réactions entre par

Improved methodology for analysis and design of Molten Salt Reactors

Rodrigo Gonzalez Gonzaga De Oliveira

Molten Salt Reactors is a class of advanced reactors that is typically characterised by the presence of a molten inorganic salt of a fissile material as fuel. At the moment, this class of reactors is considered one of the promising options in the long run of nuclear power.In the context of safety assessment of the MSFR, a task on analysis of selected transients was planned by the SAMOFAR project, to which this thesis should contribute. In addition, fluids with high melting temperature bring with them challenges regarding solidification of the fluid and possible flow blockage, to which a method was implemented during this thesis to evaluate the consequences. The open core cavity with curved shape adds another issue where the core is highly turbulent and requires the use of unstructured mesh codes for routine analysis, adding uncertainty and computational burden to the analytical workflow. Using ATARI, the tool developed during this thesis, we:1. Analyse the MSFR together with partners in support of a safety assessment task.2. Consider the impact and how to include solidification/melting in an analytical methodology.3. Evaluate options to manage/reduce uncertainties and the burden of routine use of high-fidelity codes.After a careful derivation of the conservation laws and assembling the code main algorithms, ATARI undergoes verification in a limited scope using the Stefan problem, manufactured solutions and energy balances.The nuclear cavity benchmark was performed between SAMOFAR partners showing equivalency of the codes within the scope of the benchmark. Then the participating institutions perform analysis of the MSFR for the safety assessment task, which ATARI is unable to followup with the transient analysis. The reason for failure is identified as the requirement of appropriate acceleration schemes by OpenFOAM-based multiphysics solvers, and the lack of one in ATARI.In the lack of accurate data regarding the heat exchangers, the impact of solidification is evaluated in a few simplified cases of a pipe with square cross-section and heat exchanger. For different salt compositions and different boundary conditions, the heat exchanger is found to recover from blockage if pump maintains its momentum and some cross-flow is allowed. The recovery of cases with abrupt solidification is questionable due to the required pressure head.The use of flow baffles is investigated as an option to structure the flow inside the reactor and reduce uncertainties originating from turbulence modelling while allowing the use of structured mesh codes for analysis. The concept is tested in a chloride based MSR, which retains its breed and burn mode while showing a quasi-1D flow appropriate for modelling with legacy codes.We recommend that the nuclear cavity benchmark should be expanded to include a more demanding transient case, representative of MSFR transient analysis requirements. A case is made for homogenised and economical models as used in ATARI to enable the practical use of coupling methods that promote the simultaneous solution of equations. A niche for ATARI is found as a special-mission code for analysis of challenging cases with deforming structure and flow field that justifies the use of unstructured meshes and the added computational burden. Due to design uncertainties and their nature, it is recommended to consider solidification implicitly when analysing the MSFR fuel circuit.

EPFL2021

Delayed Hydride Cracking in Irradiated and Unirradiated Zircaloy-2 Cladding

Aaron William Colldeweih

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

An Optimal Allocation between R&D and Prototype Funding: The Case of Generation IV International Forum’s Nuclear Energy Initiative

National Research Council of the National Academies, Prospective Evaluation of Applied Energy Research and Development at DOE (Phase One): A First Look Forward (2005) proposes a cost-benefit methodology to evaluate U.S. Department of Energy’s Research, Development, and Demonstration (RD&D) programs. This paper develops the methodology for nuclear energy programs. The RD&D process is analyzed in two stages with two success probabilities: (1) that the technology will transition from the R&D Stage to the Prototype Demonstration Stage, and (2) that the technology will be adopted commercially. It models discounted expected total benefits of an RD&D program as a function of the levels of funding, stage durations, the probabilities of success, and spillovers to other technologies. Project duration is an exponential function of funding. Project success is a logistic function of funding and uncertainty. Spillovers are linear functions of funding at each stage. This specification allows calculation of the marginal effects of changes in funding on discounted expected total benefits for a single technology. The paper uses this method to offer an optimal allocation of pre-prototype R&D funding in the development of the Generation IV International Forum’s advanced nuclear energy systems under a specific parameterization and funding constraint.