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Subramanian, Gopinath

Publications and source records attributed to Subramanian, Gopinath.

8 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Use of multiscale zirconium alloy deformation models in nuclear fuel behavior analysis

Accurate prediction of cladding mechanical behavior is a key aspect of modeling nuclear fuel behavior, especially for conditions of pellet-cladding interaction (PCI), reactivity-initiated accidents (RIA), and loss of coolant accidents (LOCA). Current approaches to fuel performance modeling rely on empirical constitutive models for cladding creep, growth and plastic deformation, which are limited to the materials and conditions for which the models were developed. To improve upon this approach, a microstructurally-based zirconium alloy mechanical deformation analysis capability is being developed within the United States Department of Energy Consortium for Advanced Simulation of Light Water Reactors (CASL). Specifically, the viscoplastic self-consistent (VPSC) polycrystal plasticity modeling approach, developed by Lebensohn and Tomé [1], has been coupled with the BISON engineering scale fuel performance code to represent the mechanistic material processes controlling the deformation behavior of light water reactor (LWR) cladding. A critical component of VPSC is the representation of the crystallographic nature (defect and dislocation movement) and orientation of the grains within the matrix material and the ability to account for the role of texture on deformation. A future goal is for VPSC to obtain information on reaction rate kinetics from atomistic calculations to inform the defect and dislocation behavior models described in VPSC. The multiscale modeling of cladding deformation mechanisms allowed by VPSC far exceed the functionality of typical semi-empirical constitutive models employed in nuclear fuel behavior codes to model irradiation growth and creep, thermal creep, or plasticity. This paper describes the implementation of an interface between VPSC and BISON and provides initial results utilizing the coupled functionality.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗
DOE OSTI2016

Kinetic Interaction of Uranium Vacancies and Dislocations in UO 2

Understanding how point defects and defect clusters interact with dislocations in urania is important for a number of phenomena. For example, dislocations are one (of many) trap sites in the fuel where fission gases may accumulate and ultimately nucleate fission gas bubbles. Further, some creep mechanisms are governed by the flow of point defects to dislocations. Thus, for a variety of reasons, it is important to examine how dislocations attract and accelerate the kinetics of point defects.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

A generalized force-modified potential energy surface for mechanochemical simulations

We describe the modifications that a spatially varying external load produces on a Born-Oppenheimer potential energy surface (PES) by calculating static quantities of interest. The effects of the external loads are exemplified using electronic structure calculations (at the HF/6-31G{sup ∗∗} level) of two different molecules: ethane and hexahydro-1,3,5-trinitro-s-triazine (RDX). The calculated transition states and Hessian matrices of stationary points show that spatially varying external loads shift the stationary points and modify the curvature of the PES, thereby affecting the harmonic transition rates by altering both the energy barrier as well as the prefactor. The harmonic spectra of both molecules are blueshifted with increasing compressive “pressure.” Some stationary points on the RDX-PES disappear under application of the external load, indicating the merging of an energy minimum with a saddle point.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2015

Impact of homogeneous strain on uranium vacancy diffusion in uranium dioxide

We present a detailed mechanism of, and the effect of homogeneous strains on, the migration of uranium vacancies in UO 2 . Vacancy migration pathways and barriers are identified using density functional theory and the effect of uniform strain fields are accounted for using the dipole tensor approach. We report complex migration pathways and noncubic symmetry associated with the uranium vacancy in UO 2 and show that these complexities need to be carefully accounted for to predict the correct diffusion behavior of uranium vacancies. We show that under homogeneous strain fields, only the dipole tensor of the saddle with respect to the minimum is required to correctly predict the change in the energy barrier between the strained and the unstrained case. Diffusivities are computed using kinetic Monte Carlo simulations for both neutral and fully charged state of uranium single and divacancies. We calculate the effect of strain on migration barriers in the temperature range 800–1800 K for both vacancy types. Homogeneous strains as small as 2% have a considerable effect on diffusivity of both single and divacancies of uranium, with the effect of strain being more pronounced for single vacancies than divacancies. In contrast, the response of a given defect to strain is less sensitive to changes in the charge state of the defect. Further, strain leads to anisotropies in the mobility of the vacancy and the degree of anisotropy is very sensitive to the nature of the applied strain field for strain of equal magnitude. Our results indicate that the influence of strain on vacancy diffusivity will be significantly greater when single vacancies dominate the defect structure, such as sintering, while the effects will be much less substantial under irradiation conditions where divacancies dominate.

36 MATERIALS SCIENCE↗
DOE OSTI2012

Progress report on the incorporation of lower lengthscales into polycrystal plasticity models

Neutron irradiation produces vacancies and interstitials in the crystal lattice, with concentrations far in excess to the thermal equilibrium concentrations. These defects either recombine or migrate to sinks where they get trapped. Dislocations present in the lattice preferentially absorb either vacancies or interstitials, causing them to climb (in the absence of stress) or to climb and glide (when a resolved shear stress is present). Both mechanisms induce dimensional changes, named growth and creep, respectively.

36 MATERIALS SCIENCE↗
DOE OSTI2012

Progress report on the incorporation of lower lengthscales into polycrystal plasticity models - Part 2

Creep is the time-dependent, irreversible deformation that occurs at a low rate, at stresses below the ultimate stress of the material. Non-hydrostatic stresses (i.e. shear stresses) are required for creep. The kinds of creep relevant in cladding materials are (1) thermal creep, and (2) irradiation creep. These deformation processes require elevated temperatures, and applied stresses. Both processes are mediated by the motion of dislocations. Dislocations preferentially absorb either vacancies or interstitials, causing them to climb over obstacles. In thermal creep, the higher temperature causes high mobility of point defects, leading to a significant absorption rate of point defects by dislocations, leading to dislocation motion. In irradiation creep, the high radiation flux causes a supersaturation of point defects, that in turn leads to point defect absorption, and consequently dislocation motion. Irradiation growth is defined as the irradiation induced dimensional changes that take place in the absence of applied stresses, and is caused by the partitioning of vacancies and interstitials among two or more kinds of sinks. Each kind of sink produces strain in a different crystallographic direction, such that the final deformation is anisotropic, and is volume conserved. In this phenomenon, the high radiation flux causes the supersaturation of point defects. These point defects then tend to aggregate into dislocation loops (either interstitial of vacancy). These loops then absorb more point defects, thereby growing in size, and leading to macroscopic strain. The two phenomena of creep and growth become coupled in polycrystalline materials. Growth in a single grain is opposed by stresses caused by interaction of the growing grain with its neighbors (that are also growing, but it in different directions). These stresses initiate and/or potentiate irradiation creep. In extreme cases, such as accident scenarios, thermal creep becomes important. Internal stresses in turn affect the diffusive characteristics of point defects within a grain, and consequently affecting the phenomenon of growth. In this report, we examine the coupling between creep and growth in a polycrystalline materials, and also look at the atomistic mechanisms that contribute to these phenomena.

36 MATERIALS SCIENCE↗