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Enhancing performance of fuel cladding and duct alloys is a key means of increasing fuel burnup. This project will address the failure of fuel cladding via three-dimensional cracking models. Researchers will develop a simulation code for the failure of the fuel cladding and validate the code through experiments. The objective is to develop an algorithm to determine the failure of fuel cladding in the form of three-dimensional cracking due to prolonged exposure under varying conditions of pressure, temperature, chemical environment, and irradiation. This project encompasses the following tasks: 1. Simulate 3D crack initiation and growth under instantaneous and/or fatigue loads using a new variant of the material point method (MPM); 2. Simulate debonding of the materials in the crack path using cohesive elements, considering normal and shear traction separation laws; 3. Determine the crack propagation path, considering damage of the materials incorporated in the cohesive elements to allow the energy release rate to be minimized; 4. Simulate the three-dimensional fatigue crack growth as a function of loading histories; 5. Verify the simulation code by comparing results to theoretical and numerical studies available in the literature; 6. Conduct experiments to observe the crack path and surface profile in unused fuel cladding and validate against simulation results; and 7. Expand the adaptive mesh refinement infrastructure parallel processing environment to allow adaptive mesh refinement at the 3D crack fronts and adaptive mesh merging in the wake of cracks. Fuel cladding is made of materials such as stainless steels and ferritic steels with added alloying elements, which increase stability and durability under irradiation. As fuel cladding is subjected to water, chemicals, fission gas, pressure, high temperatures, and irradiation while in service, understanding performance is essential. In the fast fuel used in advanced burner reactors, simulations of the nuclear fuels are critical to understand the burnup, and thus the fuel efficiency.
In power ramp tests, some high burn-up boiling water reactor (BWR) fuels failed by outside-in cracking. After detailed post irradiation examinations, it was deduced that fracture of radially oriented hydrides precipitated at the outer surface had resulted in the formation of incipient radial cracks that propagated towards the inner surface. Since crack propagation is a key process that dominates the outside-in failure, experimental data of outside-in cracking are obviously required. The objective of the present study is to evaluate the threshold conditions and the velocity of the outside-in cracking during a power ramp to assess the integrity of BWR fuels during a power ramp. In the experiments, by using unirradiated Zry-2 fuel cladding tubes, the outside-in cracking was examined both with and without the thermal gradient. Without the radial thermal gradient, there was no radial delayed hydride cracking in all experimental conditions examined. On the other hand, with the radial thermal gradient, the velocity of the outside-in cracking was significantly higher than that expected from the data obtained under the isothermal condition in our previous study. The threshold depth of the incipient crack was found to be ∼0.1 mm at the hoop stress of 300 MPa. It was suggested that the outside-in cracking during the power ramp is strongly dependent on the distribution of dissolved hydrogen as a result of the thermal diffusion.
The authors of this text aim to educate the reader on nuclear power and its future potential. It focuses on nuclear accidents such as Chernobyl and Three Mile Island, and their consequences, with the understanding that there are safety lessons to be learned if nuclear power generation is going to be expanded to meet our growing energy needs.
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
It is well known that oxide fuels crack during the first rise to power, with continued fracture occurring during steady operation and especially during power ramps or accidental transients. Fractures have a very strong influence on the stress state in the fuel which, in turn, drives critical phenomena such as fission gas release, fuel creep, and eventual fuel/clad mechanical interaction. Recently, interest has been expressed in discrete fracture methods, such as the cohesive zone approach. Such models are attractive from a mechanistic and physical standpoint, since they reflect the localized nature of cracking. The precise locations where fractures initiate, as well as the crack evolution characteristics, are determined as part of the solution. This paper explores the use of finite element cohesive zone concepts to predict dynamic crack behavior in oxide fuel pellets during power-up, steady operation, and power ramping. The aim of this work is first to provide an assessment of cohesive zone models for application to fuel cracking and explore important numerical issues associated with this fracture approach. A further objective is to provide basic insight into where and when cracks form, how they interact, and how cracking effects the stress field in a fuel pellet. The ABAQUS commercial finite element code, which includes powerful cohesive zone capabilities, was used for this study. Fully-coupled thermo-mechanical behavior is employed, including the effects of thermal expansion, swelling due to solid and gaseous fission products, and thermal creep. Crack initiation is determined by a temperature-dependent maximum stress criterion, based on measured fracture strengths for UO2. Damage evolution is governed by a traction-separation relation, calibrated to data from temperature and burn-up dependent fracture toughness measurements. Numerical models are first developed in 2D based on both axisymmetric (to explore axial cracking) and plane strain (to explore radial cracking) assumptions. A 3D model is then developed, permitting simultaneous radial and axial fractures. Although fuel cracking is clearly three-dimensional, 2D models are of interest since they are simpler to implement, are much less computationally intensive, and have potential application to existing 2D fuel performance codes. Numerical issues related to cohesive zone models, such as mesh dependency and viscous stabilization, are addressed. Model results indicate that for typical oxide fuel properties, both axial and radial cracking occurs during initial heat-up, well before steady-state thermal gradients are established in the pellet. Cracking results in local stress relief and a shift in peak stress locations, leading to the initiation of new cracks. Continued growth of existing cracks, plus the initiation and growth of additional fractures, is observed during steady operation and power ramping. 3D models provide considerable insight into the progressive interactions between radial and axial cracking. Parametric studies demonstrate the effects of temperature dependent material properties on crack initiation and progression. Increasing fracture strength and toughness with temperature, leads to crack arrest in high temperature regions near the pellet's symmetry axis.