This research characterized, by atomic level molecular dynamics (MD) simulations, the early stage aggregation of the fission product xenon in a model uranium oxide nuclear fuel matrix that contributes to the long term bubble formation mechanisms that compromise the efficiency and lifetime of nuclear fuels. These atomic level computer simulations disclosed a previously unknown pre-coarsening phenomenon in which the clustering of xenon is driven by vacancy diffusion. These key findings are to be included in higher level simulations of nuclear fuel thermo-mechanical processes that model irradiation effects within nuclear fuels. Previous multiscale models did not include these early time pre-coarsened clusters that occur in the nanosecond regime and therefore will lead to a better understanding of bubble formation in nuclear fuels.
A nuclear reactor operates in an environment where complex multi-physics and multi-scale phenomena exist, and it requires consideration of coupling among neutronics, thermal hydraulics, fuel performance, chemical dynamics, and coupling between the reactor core and first circuit. Safe, reliable, and economical operation can be achieved by leveraging high-fidelity numerical simulation, and proper considerations for coupling among different physics and required to provide powerful numerical simulation tools. In the past simplistic models for some of the physics phenomena are used, with the recent development of advanced numerical methods, software design, and high-performance computing power, the appeal of multi-physics and multi-scale modeling and simulation has been broadened.
The ''Workshop on Advanced Computational Materials Science: Application to Fusion and Generation IV Fission Reactors'' was convened to determine the degree to which an increased effort in modeling and simulation could help bridge the gap between the data that is needed to support the implementation of these advanced nuclear technologies and the data that can be obtained in available experimental facilities. The need to develop materials capable of performing in the severe operating environments expected in fusion and fission (Generation IV) reactors represents a significant challenge in materials science. There is a range of potential Gen-IV fission reactor design concepts and each concept has its own unique demands. Improved economic performance is a major goal of the Gen-IV designs. As a result, most designs call for significantly higher operating temperatures than the current generation of LWRs to obtain higher thermal efficiency. In many cases, the desired operating temperatures rule out the use of the structural alloys employed today. The very high operating temperature (up to 1000 C) associated with the NGNP is a prime example of an attractive new system that will require the development of new structural materials. Fusion power plants represent an even greater challenge to structural materials development and application. The operating temperatures, neutron exposure levels and thermo-mechanical stresses are comparable to or greater than those for proposed Gen-IV fission reactors. In addition, the transmutation products created in the structural materials by the high energy neutrons produced in the DT plasma can profoundly influence the microstructural evolution and mechanical behavior of these materials. Although the workshop addressed issues relevant to both Gen-IV and fusion reactor materials, much of the discussion focused on fusion; the same focus is reflected in this report. Most of the physical models and computational methods presented during the workshop apply equally to both types of nuclear energy systems. The primary factor that differentiates the materials development path for the two systems is that nearly prototypical irradiation environments for Gen-IV materials can be found or built in existing fission reactors. This is not the case for fusion. The only fusion-relevant, 14 MeV neutron sources ever built (such as the rotating target neutron sources, RTNS-I and -II at LLNL) were relatively low-power accelerator based systems. The RTNS-II ''high'' flux irradiation volume was quite small, less than 1 cm{sup 3}, and only low doses could be achieved. The maximum dose data obtained was much less than 0.1 dpa. Thus, RTNS-II, which last operated in 1986, provided only a limited opportunity for fundamental investigations of the effects of 14 MeV neutrons characteristic of DT fusion.
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
Because of various excellent properties, SiC has been proposed for many applications in nuclear reactors including cladding layers in fuel rod, fission products container in TRISO fuel, and first wall/blanket in magnetic controlled fusion reactors. Upon exposure to high energy radiation environments, point defects and defect clusters are generated in materials in amounts significantly exceeding their equilibrium concentrations. The accumulation of defects can lead to undesired consequences such as crystalline-to-amorphous transformation1, swelling, and embrittlement, and these phenomena can adversely affect the lifetime of SiC based components in nuclear reactors. It is of great importance to understand the accumulation process of these defects in order to estimate change in properties of this material and to design components with superior ability to withstand radiation damages. Defect clusters are widely in SiC irradiated at the operation temperatures of various reactors. These clusters are believed to cause more than half of the overall swelling of irradiated SiC and can potentially lead to lowered thermal conductivity and mechanical strength. It is critical to understand the formation and growth of these clusters. Diffusion of these clusters is one importance piece to determine the growth rate of clusters; however it is unclear so far due to the challenges in simulating rare events. Using a combination of kinetic Activation Relaxation Technique with empirical potential and ab initio based climbing image nudged elastic band method, I performed an extensive search of the migration paths of the most stable carbon tri-interstitial cluster in SiC. This research reveals paths with the lowest energy barriers to migration, rotation, and dissociation of the most stable cluster. Based on these energy barriers, I concluded defect clusters are thermally immobile at temperatures lower than 1500 K and can dissociate into smaller clusters and single interstitials at temperatures beyond that. Even though clusters cannot diffuse by thermal vibrations, we found they can migrate at room temperature under the influence of electron radiation. This is the first direct observation of radiation-induced diffusion of defect clusters in bulk materials. We show that the underlying mechanism of this athermal diffusion is elastic collision between incoming electrons and cluster atoms. Our findings suggest that defect clusters may be mobile under certain irradiation conditions, changing current understanding of cluster annealing process in irradiated SiC. With the knowledge of cluster diffusion in SiC demonstrated in this thesis, we now become able to predict cluster evolution in SiC with good agreement with experimental measurements. This ability can enable us to estimate changes in many properties of irradiated SiC relevant for its applications in reactors. Internal interfaces such as grain boundaries can behave as sinks to radiation induced defects. The ability of GBs to absorb, transport, and annihilate radiation-induced defects (sink strength) is important to understand radiation response of polycrystalline materials and to better design interfaces for improved resistance to radiation damage. Nowadays, it is established GBs' sink strength is not a static property but rather evolves with many factors, including radiation environments, grain size, and GB microstructure. In this thesis, I investigated the response of small-angle tilt and twist GBs to point defects fluxes in SiC. First of all, I found the pipe diffusion of interstitials in tilt GBs is slower than bulk diffusion. This is because the increased interatomic distance at dislocation cores raises the migration barrier of interstitial dumbbells. Furthermore, I show that both the annihilation of interstitials at jogs and jog nucleation from clusters are diffusion-controlled and can occur under off-stoichiometric interstitial fluxes. Finally, a dislocation line model is developed to predict the role of tilt GBs in annihilating radiation damage. The model predicts the role of tilt GBs in annihilating defects depends on the rate of defects segregation to and diffusion along tilt GBs. Tilt GBs mainly serve as diffusion channel for defects to reach other sinks when defect diffusivity is high at boundaries. When defect diffusivity is low, most of the defects segregated to tilt GBs are annihilated by dislocation climb. Up-to-date, the response of twist GBs under irradiation has been rarely reported in literature and is still unclear. It is important to develop atom scale insight on this question in order to predict twist GBs' sink strength for a better understanding of radiation response of polycrystalline materials. By using a combination of molecular dynamics and grand canonical Monte Carlo, here I demonstrate the defect kinetics in {001} and {111} twist GBs and the microstructural evolution of these GBs under defect fluxes in SiC. I found due to the deep potential well for interstitials at dislocation intersections within the interface, the mobility of defects on dislocation grid is retard and this leads to defect accumulation at GBs for many cases. Furthermore, I conclude both types of twist GBs have to form mixed dislocations with edge component in order to absorb accumulated interstitials at the interface. The formation of mixed dislocation is either by interstitial loop nucleation or by dislocation reactions at the interface. The continuous formation and climb of these mixed dislocations make twist GBs unsaturatable sinks to radiation induced defects.
Presenting papers from the 2013 annual meeting of The Minerals, Metals & Materials Society (TMS), this volume covers developments in all aspects of high temperature electrochemistry, from the fundamental to the empirical and from the theoretical to the applied.
To meet the demands of exponentially growing global energy consumption, extensive research is necessary for future fusion and Generation IV fission reactor power plants; however, structural material performance is a limiting factor. Conventional materials cannot withstand the complex thermomechanical loading, higher operating temperatures, and high irradiation doses. Consequently, research efforts are underway to develop and qualify novel structural materials. Likewise, research on fuel materials seeks to understand and mitigate issues such as oxidation and evolution of the fuel microstructure to high burn-up. To minimize trial-and-error experiments, major fission and fusion materials programs have expanded their use of multiscale materials modeling and simulation. The modeling efforts, in concert with specific experimental validation activities, complement the large experimental programs. This book highlights nuclear materials issues and challenges on experimental and modeling platforms. Topics include: fundamental defect behavior; experimental and theoretical investigations of radiation damage; behavior of oxide dispersion strengthened steels and other advanced structural materials; corrosion and surface coatings; and modeling and performance of nuclear fuels.
High-performance alloys that can withstand operation in hazardous nuclear environments are critical to presentday in-service reactor support and maintenance and are foundational for reactor concepts of the future. With commercial nuclear energy vendors and operators facing the retirement of staff during the coming decades, much of the scholarly knowledge of nuclear materials pursuant to appropriate, impactful, and safe usage is at risk. Led by the multi-award winning editorial team of G. Robert Odette (UCSB) and Steven J. Zinkle (UTK/ORNL) and with contributions from leaders of each alloy discipline, Structural Alloys for Nuclear Energy Applications aids the next generation of researchers and industry staff developing and maintaining steels, nickel-base alloys, zirconium alloys, and other structural alloys in nuclear energy applications. This authoritative reference is a critical acquisition for institutions and individuals seeking state-of-the-art knowledge aided by the editors' unique personal insight from decades of frontline research, engineering and management. - Focuses on in-service irradiation, thermal, mechanical, and chemical performance capabilities. - Covers the use of steels and other structural alloys in current fission technology, leading edge Generation-IV fission reactors, and future fusion power reactors. - Provides a critical and comprehensive review of the state-of-the-art experimental knowledge base of reactor materials, for applications ranging from engineering safety and lifetime assessments to supporting the development of advanced computational models.
