Travis W. Knight

Processing and fabrication of mixed uranium/refractory metal carbide fuels with liquid-phase sintering

Abstract Optimization of powder processing techniques were sought for the fabrication of single-phase, solid-solution mixed uranium/refractory metal carbide nuclear fuels – namely (U, Zr, Nb)C. These advanced, ultra-high temperature nuclear fuels have great potential for improved performance over graphite matrix, dispersed fuels tested in the Rover/NERVA program of the 1960s and early 1970s. Hypostoichiometric fuel samples with carbon-to-metal ratios of 0.98, uranium metal mole fractions of 5% and 10%, and porosities less than 5% were fabricated. These qualities should provide for the longest life and highest performance capability for these fuels. Study and optimization of processing methods were necessary to provide the quality assurance of samples for meaningful testing and assessment of performance for nuclear thermal propulsion applications. The processing parameters and benefits of enhanced sintering by uranium carbide liquid-phase sintering were established for the rapid and effective consolidation and formation of a solid-solution mixed carbide nuclear fuel.

Enhanced Thermal Conductivity for LWR Fuel

Abstract The thermal conductivity of the fuel in today’s light water reactors, uranium dioxide (UO2), can be improved by incorporating a uniformly distributed heat-conducting network of a higher-conductivity material: silicon carbide (SiC). The higher thermal conductivity of SiC along with its other prominent reactor-grade properties makes it a potential material to address some of the related issues when used in UO2 (97% theoretical density). This ongoing research, in collaboration with the University of Florida, aims to investigate the feasibility and development of a formal methodology for producing the resultant composite oxide fuel. Calculations of the effective thermal conductivity (ETC) of the new fuel as a function of percent SiC for certain percentages and as a function of temperature are presented as a preliminary approach. The ETCs are obtained at different temperatures from 600 to 1600 K. The corresponding polynomial equations for the temperature-dependent thermal conductivities are given based on the simulation results. The heat transfer mechanism in this fuel is explained using a finite volume approach and validated against existing empirical models. FLUENT 6.1.22 was used for the thermal conductivity calculations and to estimate the reduction in centerline temperatures achievable within such a fuel rod. Later, the computer codes COMBINE-PC and VENTURE-PC were employed to estimate the fuel enrichment required to maintain the same burnup levels corresponding to a volume percent addition of SiC.

Virtual radiation fields : A virtual environment tool for radiological analysis and simulation

A virtual reality system was developed for computational and graphical modeling and simulation of radiation environments. This system, called Virtual Radiation Fields (VRF), demonstrates the usefulness of radiological analysis in simulation-based design for predicting radiation doses for robotic equipment and personnel working in a radiation environment. The system was developed for use in determining the radiation doses for robotic equipment to be used in tank-waste retrieval operations at the Hanford National Laboratory. As a reference case, specific application is made to simulate cleanup operations for Hanford tank C-106. A three-dimensional model representation of the tank and its predicted radiation levels are presented and analyzed. Tank cleanup operations were simulated to understand how radiation levels change during the cleanup phase and to predict cumulative radiation doses to robotic equipment to aid in the development of maintenance and replacement schedules.

Processing of pseudo-ternary carbide fuels for high temperature space nuclear reactors

Single phase, solid-solution pseudo-ternary carbides such as (U, Zr, Nb)C, hold significant promise for space nuclear power and propulsion applications because of their high melting points (typically greater than 3200 K), thermochemical stability in a hot hydrogen environment, and high thermal conductivity. Their projected endurance at very high temperatures far exceeds that of fuels previously tested and signifies their potential as a fuel for increased performance characteristics (i.e. higher specific impulse and/or longer lifetime, etc.). However, insufficient test data exist under nuclear thermal propulsion (NTP) conditions of temperature and hot hydrogen environment to fully evaluate their performance. An investigation into processing techniques was conducted in order to produce a series of pseudo-ternary carbide samples for characterization and testing. Consideration was given to the real world challenges of manufacturing full-scale fuel elements. Particular consideration was given to the fabrication requirements for the innovative, square-lattice honeycomb (SLHC) fuel elements for advanced NTP cores. This paper outlines the background and technical considerations important to mixed carbide nuclear fuel development and describes the preliminary results in developing processing techniques for pseudo-ternary carbide nuclear fuels.

Potential Impacts of Modeling Full Reactor Cores Using Combined Fuel Performance and Thermal Hydraulics Codes

Abstract Nuclear reactor systems codes have the ability to model the system response in an accident scenario based on known initial conditions (ICs) at the onset of the transient. However, there has been a tendency for these codes to lack the detailed thermomechanical fuel rod response models needed for best-estimate prediction of fuel rod failure. Alternatively, the reverse can be said about fuel performance codes; they can lack the ability to capture and model the thermal-hydraulic (T-H) influence of adjacent fuel rods and the rod’s location in the reactor core. This work analyzes the limitations in using fuel performance codes to represent in-reactor conditions as determined by full-core T-H codes. The codes used in this analysis are the U.S. Nuclear Regulatory Commission’s steady-state fuel performance code FRAPCON-3.5 and T-H code TRACE-V5P3. In order to assess the impact of the limitations found in the codes, several modifications were made to all of the codes to improve code-to-code consistency. The modifications to the fuel performance code include adding the ability to model gamma-ray heating and providing realistic core coolant conditions. The T-H code modifications include adding the ability to model the fuel with axially varying burnup-dependent fuel and cladding dimensional changes and corrosion characteristics. The fuel in a Westinghouse four-loop pressurized water reactor was modeled to assess the impacts these modifications have on fuel performance and ICs for transient analysis. The results of this study show that current modeling assumptions (and limitations) can yield both conservative and nonconservative results on several important licensing criteria.

Production of High-Density Uranium Carbide Compacts for Use in Composite Nuclear Fuels

Abstract Composite nuclear fuel consisting of uranium carbide (UC) fuel microspheres dispersed in an inert matrix is one of the fuel forms being actively considered for use in gas-cooled fast reactors (GFRs). High-density UC electrodes were required for the production of fuel microspheres by the rotating electrode method as an alternate method to the sol-gel particle production route. These compacts (to serve as electrodes) were fabricated by the exothermic combustion synthesis reaction of uranium hydride and graphite powders. Ignition of combustion synthesis was then followed by solid-state sintering at different temperatures of 1521, 1779, and 1929°C. During the course of testing the electrodes for microsphere production, it was found that the structural integrity of the electrodes and thus their suitability for microsphere production depended on the microstructural characteristics of the electrodes. Those produced at higher temperatures (1929°C) had higher densities (86.6% theoretical density) and lower open porosities (2.3%) and were shown to withstand the mechanical forces and thermal stresses imposed by this microsphere production method. The processing conditions were chosen to evaluate sintering characteristics of UC and to the extent possible to find the lowest possible process temperature. Here it is understood that the intended future GFR fuel form should involve recycled fuels including minor actinides (MAs). Concern over MA volatility in high-temperature processes thus motivated investigating the effects of lower processing temperatures. It was deduced from this study that a delicate balance exists between the processing parameters, the microstructural characteristics of the electrodes, and microsphere production.

Virtual radiation fields for ALARA determination

As computing power has increased, so too has the ability to model and simulate complex systems and processes. In addition, virtual reality technology has made it possible to visualize and understand many complex scientific and engineering problems. For this reason, a virtual dosimetry program called Virtual Radiation Fields (VRF) is developed to model radiation dose rate and cumulative dose to a receptor operating in a virtual radiation environment. With the design and testing of many facilities and products taking place in the virtual world, this program facilitates the concurrent consideration of radiological concerns during the design process. Three-dimensional (3D) graphical presentation of the radiation environment is made possible through the use of IGRIP, a graphical modeling program developed by Deneb Robotics, Inc. The VRF simulation program was designed to model and display a virtual dosimeter. As a demonstration of the program`s capability, the Hanford tank, C-106, was modeled to predict radiation doses to robotic equipment used to remove radioactive waste from the tank. To validate VRF dose predictions, comparison was made with reported values for tank C-106, which showed agreement to within 0.5%. Graphical information is presented regarding the 3D dose rate variation inside the tank. Cumulative dose predictions were made for the cleanup operations of tank C-106. A four-dimensional dose rate map generated by VRF was used to model the dose rate not only in 3D space but also as a function of the amount of waste remaining in the tank. This allowed VRF to predict dose rate at any stage in the waste removal process for an accurate simulation of the radiological conditions throughout the tank cleanup procedure.

Lattice Parameter Behavior with Different Nd and O Concentrations in (U1−yNdy)O2±X Solid Solution

Abstract The solid solution of (U1−yFPy)O2±x has the same fluorite structure as UO2±x, and the lattice parameter is affected by dissolved fission product and oxygen concentrations. The relation between the lattice parameter and the concentrations of neodymium and oxygen in the fluorite structure of (U1−yNdy)O2±x was investigated using X-ray diffraction. The lattice parameter behavior in the (U1−yNdy)O2±x solid solution shows a linear change as a function of the oxygen-to-metal ratio and solubility of neodymium. The lattice parameter depends on the radii of ions forming the fluorite structure and also can be expressed by a particular rule (modified Vegard’s law). The numerical analyses of the lattice parameters for the stoichiometric and nonstoichiometric solid solutions were conducted, and the lattice parameter model for the (U1−yNdy)O2±x solid solution was assessed. A very linear relationship between the lattice parameter and the Nd and O concentration for the stoichiometry and nonstoichiometry of the (U1−yNdy)O2±x solid solution was verified.

Status of advanced carbide fuels: Past, present, and future

Solid solution, mixed uranium/refractory metal carbide fuels such as (U, Zr, Nb)C, so called ternary carbide or tri-carbide fuels have great potential for applications in next generation advanced nuclear power reactors. Because of their high melting points, high thermal conductivity, improved resistance to hot hydrogen corrosion, and good fission product retention, these advanced nuclear fuels have great potential for high performance reactors with increased safety margins. Despite these many benefits, some concerns regarding carbide fuels include compatibility issues with coolant and/or cladding materials and their endurance under the extreme conditions associated with nuclear thermal propulsion. The status of these fuels is reviewed to characterize their performance for space nuclear power applications. Results of current investigations are presented and as well as future directions of study for these advanced nuclear fuels.

Square lattice honeycomb tri-carbide fuels for 50 to 250 KN variable thrust NTP design

Ultrahigh temperature solid solution of tri-carbide fuels are used to design an ultracompact nuclear thermal rocket generating 950 seconds of specific impulse with scalable thrust level in range of 50 to 250 kilo Newtons. Solid solutions of tri-carbide nuclear fuels such as uranium-zirconium-niobium carbide. UZrNbC, are processed to contain certain mixing ratio between uranium carbide and two stabilizing carbides. Zirconium or niobium in the tri-carbide could be replaced by tantalum or hafnium to provide higher chemical stability in hot hydrogen environment or to provide different nuclear design characteristics. Recent studies have demonstrated the chemical compatibility of tri-carbide fuels with hydrogen propellant for a few to tens of hours of operation at temperatures ranging from 2800 K to 3300 K, respectively. Fuel elements are fabricated from thin tri-carbide wafers that are grooved and locked into a square-lattice honeycomb (SLHC) shape. The hockey puck shaped SLHC fuel elements are stacked up in a...