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  1. High Fidelity Ion Beam Simulation of High Dose Neutron Irradiation

    The objective of this proposal is to demonstrate the capability to predict the evolution of microstructure and properties of structural materials in-reactor and at high doses, using ion irradiation as a surrogate for reactor irradiations. “Properties” includes both physical properties (irradiated microstructure) and the mechanical properties of the material. Demonstration of the capability to predict properties has two components. One is ion irradiation of a set of alloys to yield an irradiated microstructure and corresponding mechanical behavior that are substantially the same as results from neutron exposure in the appropriate reactor environment. Second is the capability to predict the irradiatedmore » microstructure and corresponding mechanical behavior on the basis of improved models, validated against both ion and reactor irradiations and verified against ion irradiations. Taken together, achievement of these objectives will yield an enhanced capability for simulating the behavior of materials in reactor irradiations.« less
  2. Grouping techniques for large-scale cluster dynamics simulations of reaction diffusion processes

  3. Preliminary Analysis of SiC BWR Channel Box Performance under Normal Operation

    SiC-SiC composites are being considered for applications in the core components, including BWR channel box and fuel rod cladding, of light water reactors to improve accident tolerance. In the extreme nuclear reactor environment, core components like the BWR channel box will be exposed to neutron damage and a corrosive environment. To ensure reliable and safe operation of a SiC channel box, it is important to assess its deformation behavior under in-reactor conditions including the expected neutron flux and temperature distributions. In particular, this work has evaluated the effect of non-uniform dimensional changes caused by spatially varying neutron flux and temperaturesmore » on the deformation behavior of the channel box over the course of one cycle of irradiation. These analyses have been performed using the fuel performance modeling code BISON and the commercial finite element analysis code Abaqus, based on fast flux and temperature boundary conditions have been calculated using the neutronics and thermal-hydraulics codes Serpent2 and COBRA-TF, respectively. The dependence of dimensions and thermophysical properties on fast flux and temperature has been incorporated into the material models. These initial results indicate significant bowing of the channel box with a lateral displacement greater than 6.5mm. The channel box bowing behavior is time dependent, and driven by the temperature dependence of the SiC irradiation-induced swelling and the neutron flux/fluence gradients. The bowing behavior gradually recovers during the course of the operating cycle as the swelling of the SiC-SiC material saturates. However, the bending relaxation due to temperature gradients does not fully recover and residual bending remains after the swelling saturates in the entire channel box.« less
  4. OBJECT KINETIC MONTE CARLO SIMULATIONS OF RADIATION DAMAGE ACCUMULATION IN TUNGSTEN

    The objective of this work is to understand the accumulation of radiation damage created by primary knock-on atoms (PKAs) of various energies, at 300 K and for a dose rate of 10-4 dpa/s in bulk tungsten using the object kinetic Monte Carlo (OKMC) method.
  5. Development of Multiscale Materials Modeling Techniques and Coarse- Graining Strategies for Predicting Materials Degradation in Extreme Irradiation Environments

    Exposure of metallic structural materials to irradiation environments results in significant microstructural evolution, property changes and performance degradation, which limits the extended operation of current generation light water reactors and restricts the design of advanced fission and fusion reactors [1-8]. This effect of irradiation on materials microstructure and properties is a classic example of an inherently multiscale phenomenon, as schematically illustrated in Figure 1a. Pertinent processes range from the atomic nucleus to structural component length scales, spanning more than 15 orders of magnitude. Time scales bridge more than 22 orders of magnitude, with the shortest being less than a femtosecondmore » [1,8]. Further, the mix of radiation-induced features formed and the corresponding property degradation depend on a wide range of material and irradiation variables. This emphasizes the importance of closely integrating models with high-resolution experimental characterization of the evolving radiation- damaged microstructure, including measurements performed in-situ during irradiation. In this article, we review some recent successes through the use of closely coordinated modeling and experimental studies of the defect cluster evolution in irradiated body-centered cubic materials, followed by a discussion of outstanding challenges still to be addressed, which are necessary for the development of comprehensive models of radiation effects in structural materials.« less
  6. BISON Fuel Performance Analysis of IFA-796 Rod 3 & 4 and Investigation of the Impact of Fuel Creep

    In order to improve the accident tolerance of light water reactor (LWR) fuel, alternative cladding materials have been proposed to replace the currently used zirconium (Zr)-based alloys. Of these materials, there is a particular focus on iron-chromiumaluminum (FeCrAl) alloys because they exhibit slower oxidation kinetics in high-temperature steam than Zr-alloys. This should decrease the energy release due to oxidation and slow cladding consumption in the presence of high temperature steam. These alloys should also exhibit increased “coping time” in the event of an accident scenario by improving the mechanical performance at high temperatures, allowing greater flexibility to achieve core cooling.more » As a continuation of the development of these alloys, in-reactor irradiation testing of FeCrAl cladded fuel rods has started. In order to provide insight on the possible behavior of these fuel rods as they undergo irradiation in the Halden Boiling Water Reactor, engineering analysis has been performed using FeCrAl material models implemented into the BISON fuel performance code. This milestone report provides an update on the ongoing development of modeling capability to predict FeCrAl cladding fuel performance and to provide an early look at the possible behavior of planned in-reactor FeCrAl cladding experiments. In particular, this report consists of two separate analyses. The first analysis consists of fuel performance simulations of IFA-796 rod 4 and two segments of rod 3. These simulations utilize previously implemented material models for the C35M FeCrAl alloy and UO2 to provide a bounding behavior analysis corresponding to variation of the initial fuel cladding gap thickness within the fuel rod. The second analysis is an assessment of the fuel and cladding stress states after modification of the fuel creep model that is currently implemented in the BISON fuel performance code. Effects from modifying the fuel creep model were identified for the BISON simulations of the IFA-796 rod 4 experiment, but show that varying the creep model (within the range investigated here) only provide a minimal increase in the fuel radius and maximum cladding hoop stress. Continued investigation of fuel behavioral models will include benchmarking the modified fuel creep model against available experimental data, as well as an investigation of the role that fuel cracking will play in the compliance of the fuel. Correctly calculating stress evolution in the fuel is key to assessing fuel behavior up to gap closure and the subsequent deformation of the cladding due to PCMI. The inclusion of frictional contact should also be investigated to determine the axial elongation of the fuel rods for comparison with data from this experiment.« less
  7. Recent advances in modeling and simulation of the exposure and response of tungsten to fusion energy conditions

    Under the anticipated operating conditions for demonstration magnetic fusion reactors beyond ITER, structural materials will be exposed to unprecedented conditions of irradiation, heat flux, and temperature. While such extreme environments remain inaccessible experimentally, computational modeling and simulation can provide qualitative and quantitative insights into materials response and complement the available experimental measurements with carefully validated predictions. For plasma facing components such as the first wall and the divertor, tungsten (W) has been selected as the best candidate material due to its superior high-temperature and irradiation properties. In this paper we provide a review of recent efforts in computational modeling ofmore » W both as a plasma-facing material exposed to He deposition as well as a bulk structural material subjected to fast neutron irradiation. We use a multiscale modeling approach –commonly used as the materials modeling paradigm– to define the outline of the paper and highlight recent advances using several classes of techniques and their interconnection. We highlight several of the most salient findings obtained via computational modeling and point out a number of remaining challenges and future research directions« less
  8. Defect microstructural equivalence in molybdenum under different irradiation conditions at low temperatures and low doses

  9. Modeling investigation of the stability and irradiation-induced evolution of nanoscale precipitates in advanced structural materials

    Materials used in extremely hostile environment such as nuclear reactors are subject to a high flux of neutron irradiation, and thus vast concentrations of vacancy and interstitial point defects are produced because of collisions of energetic neutrons with host lattice atoms. The fate of these defects depends on various reaction mechanisms which occur immediately following the displacement cascade evolution and during the longer-time kinetically dominated evolution such as annihilation, recombination, clustering or trapping at sinks of vacancies, interstitials and their clusters. The long-range diffusional transport and evolution of point defects and self-defect clusters drive a microstructural and microchemical evolution thatmore » are known to produce degradation of mechanical properties including the creep rate, yield strength, ductility, or fracture toughness, and correspondingly affect material serviceability and lifetimes in nuclear applications. Therefore, a detailed understanding of microstructural evolution in materials at different time and length scales is of significant importance. The primary objective of this work is to utilize a hierarchical computational modeling approach i) to evaluate the potential for nanoscale precipitates to enhance point defect recombination rates and thereby the self-healing ability of advanced structural materials, and ii) to evaluate the stability and irradiation-induced evolution of such nanoscale precipitates resulting from enhanced point defect transport to and annihilation at precipitate interfaces. This project will utilize, and as necessary develop, computational materials modeling techniques within a hierarchical computational modeling approach, principally including molecular dynamics, kinetic Monte Carlo and spatially-dependent cluster dynamics modeling, to identify and understand the most important physical processes relevant to promoting the “selfhealing” or radiation resistance in advanced materials containing nanoscale precipitates. In particular, the interfacial structure of embedded nanoscale precipitates will be evaluated by electronic- and atomic-scale modeling methods, and the efficiency of the validated interfaces for trapping point defects will next be evaluated by atomic-scale modeling (e.g., determining the sink strength of the precipitates), addressing key questions related to the optimal interface characteristics to attract point defects and enhance their recombination. Kinetic models will also be developed to simulate microstructural evolution of the nanoscale features and irradiation produced defect clusters, and compared with observed microstructural changes.« less
  10. First-principles study of vacancy, interstitial, noble gas atom interstitial and vacancy clusters in bcc-W

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