Engineering topics

Tonks, M. R.

Publications and source records attributed to Tonks, M. R..

4 records

DOE OSTI2019

The grain-size effect on thermal conductivity of uranium dioxide

We have investigated the grain-boundary scattering effect on the thermal transport behavior of uranium dioxide (UO 2 ). The polycrystalline samples having different grain-sizes (0.125, 1.8, and 7:2 µm) have been prepared by a spark plasma sintering technique and characterized by x-ray powder diffraction, scanning electron microscope, and Raman spectroscopy. The thermal transport properties (the thermal conductivity and thermoelectric power) have been measured in the temperature range of 2–300 K, and the results were analyzed in terms of various physical parameters contributing to thermal conductivity in these materials in relation to grain-size. We show that thermal conductivity decreases systematically with lowering grain-size in the temperatures below 30 K, where the boundary scattering dominates the thermal transport. At higher temperatures, more scattering processes are involved in the heat transport in these materials, making the analysis difficult. We determined the grain-boundary Kapitza resistance that would result in the observed increase in thermal conductivity with grain-size and compared the value with Kapitza resistances calculated for UO 2 using molecular dynamics from the literature.

36 MATERIALS SCIENCE↗

DOE OSTI2018

Milestone Report: Grain growth and fission gas behaviour in doped UO 2 M2MS-18LA0201034

Here, the microstructure of UO 2 can be modified during nuclear fuel fabrication by using additives. Several dopants (e.g. Cr, Ti, V, Mg, Nb) are used to enhance grain growth and densification during sintering. In previous work, we used atomic scale simulation techniques to identify a common interstitial solution mechanism for a range of dopants that form positively charged defects at sintering temperatures. As a result, negatively/positively charged defect concentrations were enhanced/suppressed relative to undoped fuel under the same conditions (temperature and oxygen partial pressure). Importantly U and fission gas migration is vacancy mediated, such that the increase in the concentration uranium vacancies (which have negative charge), due to doping, causes an increase in U and fission gas diffusivity. This is important because U diffusivity is the rate limiting step in mass transport during sintering and bulk fission gas diffusivity is the underlying property that governs fission gas release. In this work, we implement analytical expressions for uranium vacancy concentrations in doped fuel (based on atomic scale results) into mesoscale and engineering scale simulations of sintering and fission gas release. It is shown that densification is dramatically enhanced at 1850 K and 1950 K due to dopant solution and that Mn-doping has a greater effect than the more widely used Cr-doping. The effect of both dopants is strongly temperature dependent and is negligible below approximately 1800 K. The effect of enlarged grain size and enhanced fission gas diffusivity has been included in BISON. Fuel temperatures under normal operating conditions are not sufficient to activate the enhanced fission gas diffusivity mechanism predicted by atomic scale calculations. Therefore, the benefits of enlarged grains for reduced fission gas release should not undermined by enhanced fission gas diffusivity for doped UO 2 .

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2015

Advanced multiphysics coupling for LWR fuel performance analysis

Even the most basic nuclear fuel analysis is a multiphysics undertaking, as a credible simulation must consider at a minimum coupled heat conduction and mechanical deformation. The need for more realistic fuel modeling under a variety of conditions invariably leads to a desire to include coupling between a more complete set of the physical phenomena influencing fuel behavior, including neutronics, thermal hydraulics, and mechanisms occurring at lower length scales. This paper covers current efforts toward coupled multiphysics LWR fuel modeling in three main areas. The first area covered in this paper concerns thermomechanical coupling. The interaction of these two physics, particularly related to the feedback effect associated with heat transfer and mechanical contact at the fuel/clad gap, provides numerous computational challenges. An outline is provided of an effective approach used to manage the nonlinearities associated with an evolving gap in BISON, a nuclear fuel performance application. A second type of multiphysics coupling described here is that of coupling neutronics with thermomechanical LWR fuel performance. DeCART, a high-fidelity core analysis program based on the method of characteristics, has been coupled to BISON. DeCART provides sub-pin level resolution of the multigroup neutron flux, with resonance treatment, during a depletion or a fast transient simulation. Two-way coupling between these codes was achieved by mapping fission rate density and fast neutron flux fields from DeCART to BISON and the temperature field from BISON to DeCART while employing a Picard iterative algorithm. Finally, the need for multiscale coupling is considered. Fission gas production and evolution significantly impact fuel performance by causing swelling, a reduction in the thermal conductivity, and fission gas release. The mechanisms involved occur at the atomistic and grain scale and are therefore not the domain of a fuel performance code. However, it is possible to use lower length scale models such as those used in the mesoscale MARMOT code to compute average properties, e.g. swelling or thermal conductivity. These may then be used by an engineering-scale model. Examples of this type of multiscale, multiphysics modeling are shown.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2015

Release 1.0 of MARMOT: A Mesoscale Fuel Performance Code

MARMOT is the mesoscale fuel performance code under development as part of the US DOE Nuclear Energy Advanced Modeling and Simulation Program. With this report, we release a new version of MARMOT, summarize its implemented capabilities, and document the ongoing development of new capabilities planned for the next release. The purpose of MARMOT is to predict the coevolution of microstructure and material properties of nuclear fuel and cladding due to stress, temperature, and irradiation damage. It accomplishes this using the phase field method coupled to solid mechanics and heat conduction. MARMOT is based on the Multiphysics Object-Oriented Simulation Environment (MOOSE), and much of its basic capability in the areas of the phase field method, mechanics, and heat conduction come directly from MOOSE modules.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗