Reactivity feedbacks of a material test research reactor fueled with various low enriched uranium dispersion fuels

The reactivity feedbacks of a material test research reactor using various low enriched uranium fuels, having same uranium density were calculated. For this purpose, the original aluminide fuel (UAlx-Al) containing 4.40 gU/cm3 of an MTR was replaced with silicide (U3Si-Al and U3Si2-Al) and oxide (U3O8-Al) dispersion fuels having the same uranium density as of the original fuel. Calculations were carried out to find the fuel temperature reactivity feedback, moderator temperature reactivity feedback, moderator density reactivity feedback and moderator void reactivity feedback. Nuclear reactor analysis codes including WIMS-D4 and CITATION were employed to carry out these calculations. It was observed that the magnitudes all the respective reactivity feedbacks from 38 °C to 50 °C and 100 °C, at the beginning of life, of all the fuels were very close to each other. The fuel temperature reactivity feedback of the U3O8-Al was about 2% more than the original UAlx-Al fuel. The magnitudes of the moderator temperature, moderator density and moderator void reactivity feedbacks of all the fuels, showed very minor variations from the original aluminide fuel.