Abstract
The presence of gallium in weapons grade plutonium has raised many questions concerning its use in light water reactor (LWR) fuel rods. The biggest concern is that the gallium will migrate down the thermal gradient in the fuel rod and deposit on the inner surface of the clad, which could cause it to fail. In order to study these effects, a fuel rod thermal simulation system (FRTSS) has been developed to recreate the shape and magnitude of the temperature profile in pressurized water reactor (PWR) fuel rods. The system uses electrically heated simulated fuel rods inside of a large, natural convection heat exchanger that uses lead-bismuth eutectic (LBE) (45 <% Pb, 55 <% Bi) as the working fluid. The simulated rods consist of small diameter electric heaters, annular pellets of depleted uranium/cerium oxide doped with approximately 10 ppm of gallium, a small helium filled gap, and generic Zircaloy IV cladding. The system is controlled through a computer-based data acquisition system that is used to record temperature data and operate the various pieces of equipment. A simple mathematical model was used to design the heat exchanger and predict the temperature profile within the simulated rods. Results from system tests indicated that the mathematical model was capable of predicting heater surface temperatures within 6.15% +/- 1.82% and clad outer surface temperatures within 1.91% +/- 4.46%. In addition, the tests also revealed that the system could accurately simulate the temperature profiles of operating PWR fuel rods. Consequently, the FRTSS provides a safe and effective means for studying gallium migration in the fuel pellets and its subsequent interactions with Zircaloy IV.
Allison, Christopher Curtis (1999). The design, construction, and testing of a nuclear fuel rod thermal simulation system to study gallium/Zircaloy interactions. Master's thesis, Texas A&M University. Available electronically from
https : / /hdl .handle .net /1969 .1 /ETD -TAMU -1999 -THESIS -A446.