Abstract: Nuclear fuel cladding materials act as protective barriers for UO₂ pellets, providing essential corrosion resistance and maintaining low neutron and gamma radiation absorption cross-sections. To enhance the accident-tolerant fuel (ATF) framework and strengthen nuclear fuel cladding protection, this study investigates chromium-coated Zirconium alloy (E110) and Silicon Carbide (SiC) as cladding materials in a PWR reactor fuel assembly. The key concern addressed is the degradation of the irradiated inner and outer cladding surface, driven by fission fragments produced during nuclear reactions, steam with high temperature, and the formation of hydrides. SiC demonstrates superior resistance to chemical attacks compared to Zircaloy, oxidizes less aggressively at high temperatures, has a higher yield strength, and shows a significantly lower creep rate and neutron capture cross-section. However, the application of a chromium coating can influence the radiation response of both materials. The study aims to determine the suitable nuclear fuel cladding between Zircaloy and Silicon Carbide with Cr coating. The performance of Zircaloy (E110) and SiC cladding materials with thin chromium coatings is modeled and assessed in this work using GEANT4 simulation software. The outcomes are verified and compared to the EpiXS software dataset to guarantee precision and dependability. For modeling radiation interactions and material deterioration, GEANT4 provides a sizable dataset. These discoveries aid in creating stronger cladding materials, which raise nuclear reactor efficiency and safety. The work also shows how GEANT4 can precisely evaluate the radiation characteristics of materials exposed to gamma and neutron radiation, providing vital data for improving cladding performance in radiation environments and raising nuclear safety requirements.

Keywords: UO₂, Accident-Tolerant Fuel (ATF), Zirconium Alloy (E110), Silicon Carbide (SiC), Chromium (Cr), Geant4, EpiXS, Pressurized Water Reactor (PWR)

Pages: 19-28

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