The goal of this thesis is to study the impact of oxygen transport on thermochemistry of nuclear fuel and pellet cladding interaction. During power ramps, nuclear fuel is exposed to high temperature gradients. It undergoes chemical and structural changes. The fuel swelling leads to a mechanical contact with the cladding causing high mechanical stresses in the cladding. Simultaneously, chemically reactive gas species are released from the hot pellet center and can interact with the cladding. The combination of these chemical and mechanical factors may lead to the cladding failure by iodine stress corrosion cracking. It has been proven that oxygen transport under high temperature gradients affects irradiated fuel thermochemistry, a phenomenon which may be of importance for stress corrosion cracking. This thesis presents 3D simulations of power ramps in pressurized water reactors with the fuel performance code ALCYONE, which is part of the computing environment PLEIADES. The code has been upgraded to couple the description of irradiated fuel thermochemistry already available with oxygen transport taking into account oxygen thermal diffusion. The impact of oxygen redistribution during a power transient on irradiated fuel thermochemistry in the fuel and on chemically reactive gas release from the fuel (consisting of I(g), I2(g), CsI(g), TeI2(g), Cs(g) and Cs2(g), mainly) is studied. The simulations show that oxygen redistribution, even if moderate in magnitude, leads to the reduction of metallic oxides (molybdenum dioxide, cesium molybdate, chromium oxide) at the fuel pellet center and consequently to the release of a much greater quantity of gaseous cesium, in agreement with post-irradiation examinations. The three-dimensional calculations of the quantities of importance for iodine stress corrosion cracking (hoop stress, hoop strain, iodine partial pressure at the clad inner wall) are then used in simulations of clad crack propagation.

The goal of this thesis is to study the impact of oxygen transport on thermochemistry of nuclear fuel and pellet cladding interaction. During power ramps, nuclear fuel is exposed to high temperature gradients. It undergoes chemical and structural changes. The fuel swelling leads to a mechanical contact with the cladding causing high mechanical stresses in the cladding. Simultaneously, chemically reactive gas species are released from the hot pellet center and can interact with the cladding. The combination of these chemical and mechanical factors may lead to the cladding failure by iodine stress corrosion cracking. It has been proven that oxygen transport under high temperature gradients affects irradiated fuel thermochemistry, a phenomenon which may be of importance for stress corrosion cracking. This thesis presents 3D simulations of power ramps in pressurized water reactors with the fuel performance code ALCYONE, which is part of the computing environment PLEIADES. The code has been upgraded to couple the description of irradiated fuel thermochemistry already available with oxygen transport taking into account oxygen thermal diffusion. The impact of oxygen redistribution during a power transient on irradiated fuel thermochemistry in the fuel and on chemically reactive gas release from the fuel (consisting of I(g), I2(g), CsI(g), TeI2(g), Cs(g) and Cs2(g), mainly) is studied. The simulations show that oxygen redistribution, even if moderate in magnitude, leads to the reduction of metallic oxides (molybdenum dioxide, cesium molybdate, chromium oxide) at the fuel pellet center and consequently to the release of a much greater quantity of gaseous cesium, in agreement with post-irradiation examinations. The three-dimensional calculations of the quantities of importance for iodine stress corrosion cracking (hoop stress, hoop strain, iodine partial pressure at the clad inner wall) are then used in simulations of clad crack propagation.

(Cont.) The FEAST-OXIDE predictions has been compared to the available FFTF, EBR-II and JOYO databases, and the agreement between the code and data was found to be satisfactory. Both metal and oxide versions of FEAST are rather superior compared to many other fuel codes in the literature. Comparing metal and oxide versions, FEAST-OXIDE has a more sophisticated fission gas release and swelling model, which is based on vacancy flow. In addition, modeling of the chemistry module of the oxide fuel requires a much more detailed analysis to estimate its impact on the thermo-mechanical behavior with a reasonable accuracy. Finally, the melting of the oxide fuel and its effect on the thermo-mechanical performance have been modeled in case of transient scenarios.

Materials in a nuclear environment are exposed to extreme conditions of radiation, temperature and/or corrosion, and in many cases the combination of these makes the material behavior very different from conventional materials. This is evident for the four major technological challenges the nuclear technology domain is facing currently: (i) long-term operation of existing Generation II nuclear power plants, (ii) the design of the next generation reactors (Generation IV), (iii) the construction of the ITER fusion reactor in Cadarache (France), (iv) and the intermediate and final disposal of nuclear waste. In order to address these challenges, engineers and designers need to know the properties of a wide variety of materials under these conditions and to understand the underlying processes affecting changes in their behavior, in order to assess their performance and to determine the limits of operation. Comprehensive Nuclear Materials, Second Edition, Seven Volume Set provides broad ranging, validated summaries of all the major topics in the field of nuclear material research for fission as well as fusion reactor systems. Attention is given to the fundamental scientific aspects of nuclear materials: fuel and structural materials for fission reactors, waste materials, and materials for fusion reactors. The articles are written at a level that allows undergraduate students to understand the material, while providing active researchers with a ready reference resource of information. Most of the chapters from the first Edition have been revised and updated and a significant number of new topics are covered in completely new material. During the ten years between the two editions, the challenge for applications of nuclear materials has been significantly impacted by world events, public awareness, and technological innovation. Materials play a key role as enablers of new technologies, and we trust that this new edition of Comprehensive Nuclear Materials has captured the key recent developments. Critically reviews the major classes and functions of materials, supporting the selection, assessment, validation and engineering of materials in extreme nuclear environments Comprehensive resource for up-to-date and authoritative information which is not always available elsewhere, even in journals Provides an in-depth treatment of materials modeling and simulation, with a specific focus on nuclear issues Serves as an excellent entry point for students and researchers new to the field

Among the most important and exciting current steps forward in geo-engineering is the development of coupled numerical models. They represent the basic physics of geo-engineering processes which can include the effects of heat, water, mechanics and chemistry. Such models provide an integrating focus for the wide range of geo-engineering disciplines. The articles within this volume were originally presented at the inaugural GeoProc conference held in Stockholm and contain a collection of unusually high quality information not available elsewhere in an edited and coherent form. This collection not only benefits from the latest theoretical developments but also applies them to a number of practical and wide ranging applications. Examples include the environmental issues around radioactive waste disposal deep in rock, and the search for new reserves of oil and gas.

As fission process heats up the fuel rods, UO2 pellets stacked on top of each other swell both radially and axially, while the surrounding Zircaloy cladding creeps down, so that the pellets eventually come into contact with the clad. This exacerbates chemical degradation of the protective cladding and high stress values may enable the formation and propagation of cracks, thus threatening the integrity of the clad. Along these lines, pellet-cladding interaction establishes itself as a major concern for fuel rod design and core operation in light water reactors. Accurately modeling fuel behavior is challenging because the mechanical contact problem strongly depends on temperature distribution and the pellet-clad coupled heat transfer problem is, in turn, affected by changes in geometry induced by body deformations and stresses generated at the contact interface. Our work focuses on active set strategies to determine the actual contact area in high-fidelity coupled physics fuel performance codes. The approach consists of two steps: in the first one, we determine the boundary region on standard finite element meshes where the contact conditions shall be enforced to prevent objects from occupying the same space. For this purpose, we developed and implemented an efficient parallel search algorithm for detecting mesh inter-penetration and vertex/mesh overlap. The second step deals with solving the mechanical equilibrium taking into account the contact conditions computed in the first step. To do so, we developed a modified version of the multi-point constraint strategy. While the original algorithm was restricted to the Jacobi preconditioned conjugate gradient method, our approach works with any Krylov solver and does not put any restriction on the type of preconditioner used. The multibody thermo-mechanical contact problem is tackled using modern numerics, with continuous finite elements and a Newton-based monolithic strategy to handle nonlinearities (the one stemming from the contact condition itself as well as the one due to the temperature-dependence of the fuel thermal conductivity, for instance) and coupling between the various physics components (gap conductance sensitive to the clad-pellet distance, thermal expansion coefficient or Young's modulus affected by temperature changes, etc.). We will provide different numerical examples for contact problems using one and multiple bodies in order to demonstrate the performance of the method. The electronic version of this dissertation is accessible from http://hdl.handle.net/1969.1/152540