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Unal, C.

Publications and source records attributed to Unal, C..

7 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Potential mechanism for lanthanide transport in metallic fuels

An experimental and modeling effort to determine the solubility and diffusivity of certain lanthanides in liquid metals has been investigated to assess the feasibility of liquid-like lanthanide transport mechanism in metallic fuels. We developed a conceptual pore model to better describe the lanthanide transport behavior through the fuel. Our results suggests that isolated pores will act most likely as lanthanide sink and the transport mechanism of lanthanides from fuel to the cladding has to occur through interconnected pores and cracks filled by sodium.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Advanced Simulation Aided Metallic Fuel Design

Nuclear engineering is reaching a domain space in which computational power and methods are starting to surpass the models that are used to simulate nuclear fuel rods. As these models are updated, simulations utilizing different compositions, operating conditions, fabrication techniques, and geometries can provide core designers with an important first step towards fuel qualification. Due to the cost associated with fuel irradiations, it can be expected that time and money spent building better fuel models may provide an attractive return during initial studies of advanced fuels, especially with regards to fuels beyond the extensively studied uranium oxide system. In general, the primary U.S. experience with irradiated fuel rods beyond traditional oxide fuel is with binary (U-Zr) and ternary (U-Pu-Zr) metal fuels. The bulk of the experimental data for metal fuels is obtained from rods irradiated in the EBR-II sodium fast reactor (SFR), thus many of the correlations developed for metal fuels comes from the post-irradiation (PIE) examination of EBR-II rods. One of the unique characteristics that set U-Pu-Zr fuels apart from traditional oxide fuels is the presence of different phases that develop in the fuel during irradiation. Each of these different phases has unique thermal, mechanical, and diffusive behavior. The transitions between these phases are not static as the initially homogenous zirconium diffuses towards the center and edge of the fuel rod, resulting in a low-Zr, high-U ring. Along with the differing material properties that occur in each phase, the increase in uranium corresponds to an increase in localized fission and power production. The direct coupling between the thermal and diffusion solves highlights one of the many examples of tightly coupled phenomenon that are present in nuclear fuel, and motivation for the development of codes that simultaneously solve each piece of physics in a fully-coupled manner.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Simulation of fission gas diffusion in UO{sub 2} by vacancy clusters

The low solubility of fission gas in nuclear fuel leads to inter- and intra-granular bubbles. These bubbles affect performance of the fuel rod and increase the fuel temperature through the increase of porosity in the fuel pellets and decreased thermal conductivity across the pellet-clad gap as gas is released into the plenum. The nucleation, growth, and inter-connectivity of the fission gas bubbles ultimately depend on the diffusivity of gas atoms through the fuel matrix. Using mechanisms and data previously established from density functional theory (DFT) and empirical potential calculations, diffusion and reaction models have been derived for xenon atoms, uranium atoms, and vacancies in the substitutional and interstitial locations. Although similar to previous models, the diffusion and reaction rates have been re-derived in a more robust manner. Previous studies have under-predicted the experimentally observed diffusivity of xenon atoms in UO{sub 2}, however it has been previously shown that by including xenon atoms bound to two vacancies, the calculated diffusivity approaches the measured data. The derived model for species diffusion and inter-species reactions has been implemented in the MARMOT phase-field code in a modular structure so as to capture the enhanced diffusion of xenon due to bound gas + vacancy-cluster movement. The MARMOT code allows for the diffusion, reaction, and source rates to be solved simultaneously, resulting in steady state or transient concentrations of each species, and ultimately the diffusion behavior of xenon atoms. By including the diffusion of clusters of different combinations of xenon and vacancies, the calculated diffusivity more closely captures the measured diffusivity of xenon when using preliminary parameters. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Developing Soret Transport Coefficients for Lanthanides from Experimental Results

Understanding the nature of lanthanide transport in metallic fuels and its relation to fuel clad chemical interaction (FCCI) has been has been a topic that is long on interest and short on data. Recent experiments have shown a strong temperature-gradient based diffusion (Soret diffusion) and have thus provided vitally important information for modelers. However, details about the actual coefficients and many of the relevant processes that crucially participate in the cerium migration and precipitation process have not yet been disclosed. Using a combination of computing tools (primarily python and BISON) we are exploring methods that may extract useful information from visual data. In this specific example, we are using what is formally called 'blob detection' algorithms to analyzed the precipitate distribution of cerium in a U-Zr metal that has had a fixed temperature gradient enforced upon it for a specified period of time. These algorithms, once properly adapted to the problem at hand, may be useful at extracting information from radial and axial scans of metallic fuels which are much more abundant than direct measurements of lanthanides. We have developed a model that qualitatively works to move mass along a temperature gradient and precipitate heavily in the cold regions. Additionally, we have observed that the solubility of cerium in U-Zr held under this kind of temperature gradient is subject to some amount of U-Zr constituent redistribution. Our results predict movement of the Zr toward the hot end of the rod followed by a U-rich layer with the cooler 3/5 of the rod staying close the the initial U-Zr distribution. A more complete analysis will define groups of correlated parameters that correctly simulate the resulting distribution of precipitates. This restricted parameters space will give smaller bounds to modelers and will expedite the analysis of future works as new and more accurate data are made available. A follow-up experiment that started with 1/10 of the initial concentration of cerium may provide valuable additional information with regard to the effect of the rates of precipitation and dissolution as well as demonstrating the potential rapidity of migration in a lower burnup scenario. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Separate effects identification via casting process modeling for experimental measurement of U–Pu–Zr alloys

Computational simulations of gravity casting processes for metallic U–Pu–Zr nuclear fuel rods have been performed using a design-of-experiments technique to determine the fluid flow, liquid heat transfer, and solid heat transfer parameters which most strongly influence the process solidification speed and fuel rod porosity. The results are used to make recommendations for the best investment of experimental time and effort to measure process parameters.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2013

Metallic Fuel Casting Development and Parameter Optimization Simulations

One of the advantages of metallic fuel is the abilility to cast the fuel slugs to near net shape with little additional processing. However, the high aspect ratio of the fuel is not ideal for casting. EBR-II fuel was cast using counter gravity injection casting (CGIC) but, concerns have been raised concerning the feasibility of this process for americium bearing alloys. The Fuel Cycle Research and Development program has begun developing gravity casting techniques suitable for fuel production. Compared to CGIC gravity casting does not require a large heel that then is recycled, does not require application of a vacuum during melting, and is conducive to re-usable molds. Development has included fabrication of two separate benchscale, approximately 300 grams, systems. To shorten development time computer simulations have been used to ensure mold and crucible designs are feasible and to identify which fluid properties most affect casting behavior and therefore require more characterization.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗