Analysis of Optimum Accident Tolerant Fuel and Cladding Behavior in Advanced Pressurized Water Reactor

Abstract

In PWR reactors, a higher temperature than the normal operation rate causes an increase in the oxidation rate between the fuel and the clad (UO 2 / Zircaloy), and this results in the release of large quantities of hydrogen, which leads to an increase in pressure and temperature inside the reactor core and also on the walls of the pressure vessel, and perhaps partial or total damage to the reactor core. This research examines the development of new types of fuel such as Uranium Nitride (UN), Uranium Silicate (U 3 Si 2 ). Also, two types of clads such as Silicon Carbide and (Fe-Cr-Al) alloy are tested. The neutronic and thermal properties of these new types have been studied, as they are characterized by the low probability of fuel interaction with cladding, as well as the presence of good neutronic and thermal properties in terms of thermal conductivity and heat capacity, which lead to an improvement in the safety margin during operation and also in the event of nuclear accidents.